Science.gov

Sample records for iron waste package

  1. Corrosion and environmental-mechanical characterization of iron-base nuclear waste package structural barrier materials. Annual report, FY 1984

    SciTech Connect

    Westerman, R.E.; Haberman, J.H.; Pitman, S.G.; Pulsipher, B.A.; Sigalla, L.A.

    1986-03-01

    Disposal of high-level nuclear waste in deep underground repositories may require the development of waste packages that will keep the radioisotopes contained for up to 1000 y. A number of iron-base materials are being considered for the structural barrier members of waste packages. Their uniform and nonuniform (pitting and intergranular) corrosion behavior and their resistance to stress-corrosion cracking in aqueous environments relevant to salt media are under study at Pacific Northwest Laboratory. The purpose of the work is to provide data for a materials degradation model that can ultimately be used to predict the effective lifetime of a waste package overpack in the actual repository environment. The corrosion behavior of the candidate materials was investigated in simulated intrusion brine (essentially NaCl) in flowing autoclave tests at 150/sup 0/C, and in combinations of intrusion/inclusion (high-Mg) brine environments in moist salt tests, also at 150/sup 0/C. Studies utilizing a /sup 60/Co irradiation facility were performed to determine the corrosion resistance of the candidate materials to products of brine radiolysis at dose rates of 2 x 10/sup 3/ and 1 x 10/sup 5/ rad/h and a temperature of 150/sup 0/C. These irradiation-corrosion tests were ''overtests,'' as the irradiation intensities employed were 10 to 1000 times as high as those expected at the surface of a thick-walled waste package. With the exception of the high general corrosion rates found in the tests using moist salt containing high-Mg brines, the ferrous materials exhibited a degree of corrosion resistance that indicates a potentially satisfactory application to waste package structural barrier members in a salt repository environment.

  2. WASTE PACKAGE TRANSPORTER DESIGN

    SciTech Connect

    D.C. Weddle; R. Novotny; J. Cron

    1998-09-23

    The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

  3. Waste disposal package

    DOEpatents

    Smith, M.J.

    1985-06-19

    This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

  4. Radioactive waste disposal package

    DOEpatents

    Lampe, Robert F.

    1986-11-04

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  5. Radioactive waste disposal package

    DOEpatents

    Lampe, Robert F.

    1986-01-01

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  6. Naval Waste Package Design Report

    SciTech Connect

    M.M. Lewis

    2004-03-15

    A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository.

  7. Tritium waste package

    DOEpatents

    Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

    1995-01-01

    A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

  8. Tritium waste package

    DOEpatents

    Rossmassler, R.; Ciebiera, L.; Tulipano, F.J.; Vinson, S.; Walters, R.T.

    1995-11-07

    A containment and waste package system for processing and shipping tritium oxide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within the outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen and oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB. 1 fig.

  9. Naval Waste Package Design Sensitivity

    SciTech Connect

    T. Schmitt

    2006-12-13

    The purpose of this calculation is to determine the sensitivity of the structural response of the Naval waste packages to varying inner cavity dimensions when subjected to a comer drop and tip-over from elevated surface. This calculation will also determine the sensitivity of the structural response of the Naval waste packages to the upper bound of the naval canister masses. The scope of this document is limited to reporting the calculation results in terms of through-wall stress intensities in the outer corrosion barrier. This calculation is intended for use in support of the preliminary design activities for the license application design of the Naval waste package. It examines the effects of small changes between the naval canister and the inner vessel, and in these dimensions, the Naval Long waste package and Naval Short waste package are similar. Therefore, only the Naval Long waste package is used in this calculation and is based on the proposed potential designs presented by the drawings and sketches in References 2.1.10 to 2.1.17 and 2.1.20. All conclusions are valid for both the Naval Long and Naval Short waste packages.

  10. Packaged Waste Treatment

    NASA Technical Reports Server (NTRS)

    1977-01-01

    This Jacksonville, Florida, apartment complex has a wastewater treatment system which clears the water, removes harmful microorganisms and reduces solid residue to ash. It is a spinoff from spacecraft waste management and environmental control technology.

  11. Seal welded cast iron nuclear waste container

    DOEpatents

    Filippi, Arthur M.; Sprecace, Richard P.

    1987-01-01

    This invention identifies methods and articles designed to circumvent metallurgical problems associated with hermetically closing an all cast iron nuclear waste package by welding. It involves welding nickel-carbon alloy inserts which are bonded to the mating plug and main body components of the package. The welding inserts might be bonded in place during casting of the package components. When the waste package closure weld is made, the most severe thermal effects of the process are restricted to the nickel-carbon insert material which is far better able to accommodate them than is cast iron. Use of nickel-carbon weld inserts should eliminate any need for pre-weld and post-weld heat treatments which are a problem to apply to nuclear waste packages. Although the waste package closure weld approach described results in a dissimilar metal combination, the relative surface area of nickel-to-iron, their electrochemical relationship, and the presence of graphite in both materials will act to prevent any galvanic corrosion problem.

  12. YUCCA MOUNTAIN WASTE PACKAGE CLOSURE SYSTEM

    SciTech Connect

    G. Housley; C. Shelton-davis; K. Skinner

    2005-08-26

    The method selected for dealing with spent nuclear fuel in the US is to seal the fuel in waste packages and then to place them in an underground repository at the Yucca Mountain Site in Nevada. This article describes the Waste Package Closure System (WPCS) currently being designed for sealing the waste packages.

  13. Generalized waste package containment model

    SciTech Connect

    Liebetrau, A.M.; Apted, M.J.

    1985-02-01

    The US Department of Energy (DOE) is developing a performance assessment strategy to demonstrate compliance with standards and technical requirements of the Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) for the permanent disposal of high-level nuclear wastes in geologic repositories. One aspect of this strategy is the development of a unified performance model of the entire geologic repository system. Details of a generalized waste package containment (WPC) model and its relationship with other components of an overall repository model are presented in this paper. The WPC model provides stochastically determined estimates of the distributions of times-to-failure of the barriers of a waste package by various corrosion mechanisms and degradation processes. The model consists of a series of modules which employ various combinations of stochastic (probabilistic) and mechanistic process models, and which are individually designed to reflect the current state of knowledge. The WPC model is designed not only to take account of various site-specific conditions and processes, but also to deal with a wide range of site, repository, and waste package configurations. 11 refs., 3 figs., 2 tabs.

  14. Safety Analysis Report for packaging (onsite) steel waste package

    SciTech Connect

    BOEHNKE, W.M.

    2000-07-13

    The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A{sub 2}s) and is a type B packaging.

  15. Safety evaluation for packaging (onsite) concrete-lined waste packaging

    SciTech Connect

    Romano, T.

    1997-09-25

    The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site.

  16. Waste Package Design Methodology Report

    SciTech Connect

    D.A. Brownson

    2001-09-28

    The objective of this report is to describe the analytical methods and processes used by the Waste Package Design Section to establish the integrity of the various waste package designs, the emplacement pallet, and the drip shield. The scope of this report shall be the methodology used in criticality, risk-informed, shielding, source term, structural, and thermal analyses. The basic features and appropriateness of the methods are illustrated, and the processes are defined whereby input values and assumptions flow through the application of those methods to obtain designs that ensure defense-in-depth as well as satisfy requirements on system performance. Such requirements include those imposed by federal regulation, from both the U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC), and those imposed by the Yucca Mountain Project to meet repository performance goals. The report is to be used, in part, to describe the waste package design methods and techniques to be used for producing input to the License Application Report.

  17. Yucca Mountain Waste Package Closure System

    SciTech Connect

    Herschel Smartt; Arthur Watkins; David Pace; Rodney Bitsoi; Eric Larsen; Timothy McJunkin; Charles Tolle

    2006-04-01

    The current disposal path for high-level waste is to place the material into secure waste packages that are inserted into a repository. The Idaho National Laboratory has been tasked with the development, design, and demonstration of the waste package closure system for the repository project. The closure system design includes welding three lids and a purge port cap, four methods of nondestructive examination, and evacuation and backfill of the waste package, all performed in a remote environment. A demonstration of the closure system will be performed with a full-scale waste package.

  18. Yucca Mountain Waste Package Closure System

    SciTech Connect

    shelton-davis; Colleen Shelton-Davis; Greg Housley

    2005-10-01

    The current disposal path for high-level waste is to place the material into secure waste packages that are inserted into a repository. The Idaho National Laboratory has been tasked with the development, design, and demonstration of the waste package closure system for the repository project. The closure system design includes welding three lids and a purge port cap, four methods of nondestructive examination, and evacuation and backfill of the waste package, all performed in a remote environment. A demonstration of the closure system will be performed with a full-scale waste package.

  19. WASTE PACKAGE DESIGN SENSITIVITY REPORT

    SciTech Connect

    P. Mecharet

    2001-03-09

    The purpose of this technical report is to present the current designs for waste packages and determine which designs will be evaluated for the Site Recommendation (SR) or Licence Application (LA), to demonstrate how the design will be shown to comply with the applicable design criteria. The evaluations to support SR or LA are based on system description document criteria. The objective is to determine those system description document criteria for which compliance is to be demonstrated for SR; and, having identified the criteria, to refer to the documents that show compliance. In addition, those system description document criteria for which compliance will be addressed for LA are identified, with a distinction made between two steps of the LA process: the LA-Construction Authorization (LA-CA) phase on one hand, and the LA-Receive and Possess (LA-R&P) phase on the other hand. The scope of this work encompasses the Waste Package Project disciplines for criticality, shielding, structural, and thermal analysis.

  20. Packaged low-level waste verification system

    SciTech Connect

    Tuite, K.; Winberg, M.R.; McIsaac, C.V.

    1995-12-31

    The Department of Energy through the National Low-Level Waste Management Program and WMG Inc. have entered into a joint development effort to design, build, and demonstrate the Packaged Low-Level Waste Verification System. Currently, states and low-level radioactive waste disposal site operators have no method to independently verify the radionuclide content of packaged low-level waste that arrives at disposal sites for disposition. At this time, the disposal site relies on the low-level waste generator shipping manifests and accompanying records to ensure that low-level waste received meets the site`s waste acceptance criteria. The subject invention provides the equipment, software, and methods to enable the independent verification of low-level waste shipping records to ensure that the site`s waste acceptance criteria are being met. The objective of the prototype system is to demonstrate a mobile system capable of independently verifying the content of packaged low-level waste.

  1. CERAMIC WASTE FORM DATA PACKAGE

    SciTech Connect

    Amoroso, J.; Marra, J.

    2014-06-13

    The purpose of this data package is to provide information about simulated crystalline waste forms that can be used to select an appropriate composition for a Cold Crucible Induction Melter (CCIM) proof of principle demonstration. Melt processing, viscosity, electrical conductivity, and thermal analysis information was collected to assess the ability of two potential candidate ceramic compositions to be processed in the Idaho National Laboratory (INL) CCIM and to guide processing parameters for the CCIM operation. Given uncertainties in the CCIM capabilities to reach certain temperatures throughout the system, one waste form designated 'Fe-MP' was designed towards enabling processing and another, designated 'CAF-5%TM-MP' was designed towards optimized microstructure. Melt processing studies confirmed both compositions could be poured from a crucible at 1600{degrees}C although the CAF-5%TM-MP composition froze before pouring was complete due to rapid crystallization (upon cooling). X-ray diffraction measurements confirmed the crystalline nature and phase assemblages of the compositions. The kinetics of melting and crystallization appeared to vary significantly between the compositions. Impedance spectroscopy results indicated the electrical conductivity is acceptable with respect to processing in the CCIM. The success of processing either ceramic composition will depend on the thermal profiles throughout the CCIM. In particular, the working temperature of the pour spout relative to the bulk melter which can approach 1700{degrees}C. The Fe-MP composition is recommended to demonstrate proof of principle for crystalline simulated waste forms considering the current configuration of INL's CCIM. If proposed modifications to the CCIM can maintain a nominal temperature of 1600{degrees}C throughout the melter, drain, and pour spout, then the CAF-5%TM-MP composition should be considered for a proof of principle demonstration.

  2. Engineered waste-package-system design specification

    SciTech Connect

    Not Available

    1983-05-01

    This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity.

  3. Waste Package Component Design Methodology Report

    SciTech Connect

    D.C. Mecham

    2004-07-12

    This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational

  4. Prevention policies addressing packaging and packaging waste: Some emerging trends.

    PubMed

    Tencati, Antonio; Pogutz, Stefano; Moda, Beatrice; Brambilla, Matteo; Cacia, Claudia

    2016-10-01

    Packaging waste is a major issue in several countries. Representing in industrialized countries around 30-35% of municipal solid waste yearly generated, this waste stream has steadily grown over the years even if, especially in Europe, specific recycling and recovery targets have been fixed. Therefore, an increasing attention starts to be devoted to prevention measures and interventions. Filling a gap in the current literature, this explorative paper is a first attempt to map the increasingly important phenomenon of prevention policies in the packaging sector. Through a theoretical sampling, 11 countries/states (7 in and 4 outside Europe) have been selected and analyzed by gathering and studying primary and secondary data. Results show evidence of three specific trends in packaging waste prevention policies: fostering the adoption of measures directed at improving packaging design and production through an extensive use of the life cycle assessment; raising the awareness of final consumers by increasing the accountability of firms; promoting collaborative efforts along the packaging supply chains.

  5. A waste package strategy for regulatory compliance

    SciTech Connect

    Stahl, D.; Cloninger, M.O.

    1990-04-01

    This paper summarizes the strategy given in the Site Characterization Plan for demonstrating compliance with the post closure performance objectives for the waste package and the Engineered Barrier System contained in the Code of Federal Regulations. The strategy consists of the development of a conservative waste package design that will meet the regulatory requirements with sufficient margin for uncertainty using a multi-barrier approach that takes advantage of the unsaturated nature of the Yucca Mountain site. 7 refs., 1 fig.

  6. Nuclear waste package materials testing report: basaltic and tuffaceous environments

    SciTech Connect

    Bradley, D.J.; Coles, D.G.; Hodges, F.N.; McVay, G.L.; Westerman, R.E.

    1983-03-01

    The disposal of high-level nuclear wastes in underground repositories in the continental United States requires the development of a waste package that will contain radionuclides for a time period commensurate with performance criteria, which may be up to 1000 years. This report addresses materials testing in support of a waste package for a basalt (Hanford, Washington) or a tuff (Nevada Test Site) repository. The materials investigated in this testing effort were: sodium and calcium bentonites and mixtures with sand or basalt as a backfill; iron and titanium-based alloys as structural barriers; and borosilicate waste glass PNL 76-68 as a waste form. The testing also incorporated site-specific rock media and ground waters: Reference Umtanum Entablature-1 basalt and reference basalt ground water, Bullfrog tuff and NTS J-13 well water. The results of the testing are discussed in four major categories: Backfill Materials: emphasizing water migration, radionuclide migration, physical property and long-term stability studies. Structural Barriers: emphasizing uniform corrosion, irradiation-corrosion, and environmental-mechanical testing. Waste Form Release Characteristics: emphasizing ground water, sample surface area/solution volume ratio, and gamma radiolysis effects. Component Compatibility: emphasizing solution/rock, glass/rock, glass/structural barrier, and glass/backfill interaction tests. This area also includes sensitivity testing to determine primary parameters to be studied, and the results of systems tests where more than two waste package components were combined during a single test.

  7. Aqueous Corrosion Rates for Waste Package Materials

    SciTech Connect

    S. Arthur

    2004-10-08

    The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports.

  8. Coupled process modeling and waste package performance

    SciTech Connect

    McGrail, B.P.; Engel, D.W.

    1992-11-01

    The interaction of borosilicate waste glasses with water has been studied extensively and reasonably good models are available that describe the reaction kinetics and solution chemical effects. Unfortunately, these models have not been utilized in performance assessment analyses, except in estimating radionuclide solubilities at the waste form surface. A geochemical model has been incorporated in the AREST code to examine the coupled processes of glass dissolution and transport within the engineering barrier system. Our calculations show that the typical assumptions used in performance assessment analyses, such as fixed solubilities or constant reaction rate at the waste form surface, do not always give conservative or realistic predictions of radionuclide release. Varying the transport properties of the waste package materials is shown to give counterintuitive effects on the release rates of some radionuclides. The use of noncoupled performance assessment models could lead a repository designer to an erroneous conclusion regarding the relative benefit of one waste package design or host rock setting over another.

  9. Horizontal Drop of 21- PWR Waste Package

    SciTech Connect

    A.K. Scheider

    2007-01-31

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design.

  10. 10 CFR 60.143 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... package monitoring program shall include laboratory experiments which focus on the internal condition of... the laboratory experiments. (d) The waste package monitoring program shall continue as long...

  11. 10 CFR 60.143 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... package monitoring program shall include laboratory experiments which focus on the internal condition of... the laboratory experiments. (d) The waste package monitoring program shall continue as long...

  12. Groundwater removal near heat dissipating waste packages

    SciTech Connect

    Manteufel, R.D.

    1996-12-31

    The thermohydrologic environment of heat-dissipating nuclear waste packages in a subsurface repository is affected by ventilation of the facility prior to permanent closure. Heat dissipated by the waste will raise the temperature of host rock and vaporize groundwater. Ventilation will remove some heat and water vapor from the subsurface, creating a desiccated region surrounding the waste packages. The resulting hot, dry environment will tend to favorably extend the containment time of the waste. This work evaluates the transient temperature field near emplacement drifts and predicts the extent of rock dryout and removal of groundwater. For two hypothetical ventilation schemes with 30-yr-old fuel and repository loading of 40 metric tons of uranium (MTU) per acre, about 4 to 5 m of rock surrounding the drifts are predicted to be dried during the preclosure period.

  13. Igneous Intrusion Impacts on Waste Packages and Waste Forms

    SciTech Connect

    P. Bernot

    2004-08-16

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

  14. WASTE PACKAGE REMEDIATION SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect

    N.D. Sudan

    2000-06-22

    The Waste Package Remediation System remediates waste packages (WPs) and disposal containers (DCs) in one of two ways: preparation of rejected DC closure welds for repair or opening of the DC/WP. DCs are brought to the Waste Package Remediation System for preparation of rejected closure welds if testing of the closure weld by the Disposal Container Handling System indicates an unacceptable, but repairable, welding flaw. DC preparation of rejected closure welds will require removal of the weld in such a way that the Disposal Container Handling System may resume and complete the closure welding process. DCs/WPs are brought to the Waste Package Remediation System for opening if the Disposal Container Handling System testing of the DC closure weld indicates an unrepairable welding flaw, or if a WP is recovered from the subsurface repository because suspected damage to the WP or failure of the WP has occurred. DC/WP opening will require cutting of the DC/WP such that a temporary seal may be installed and the waste inside the DC/WP removed by another system. The system operates in a Waste Package Remediation System hot cell located in the Waste Handling Building that has direct access to the Disposal Container Handling System. One DC/WP at a time can be handled in the hot cell. The DC/WP arrives on a transfer cart, is positioned within the cell for system operations, and exits the cell without being removed from the cart. The system includes a wide variety of remotely operated components including a manipulator with hoist and/or jib crane, viewing systems, machine tools for opening WPs, and equipment used to perform pressure and gas composition sampling. Remotely operated equipment is designed to facilitate DC/WP decontamination and hot cell equipment maintenance, and interchangeable components are provided where appropriate. The Waste Package Remediation System interfaces with the Disposal Container Handling System for the receipt and transport of WPs and DCs. The Waste

  15. IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS

    SciTech Connect

    P. Bernot

    2004-04-19

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The models are based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. The models described in this report constitute the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS:167796]) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2003 [DIRS: 166296]). The technical work plan was prepared in accordance with AP-2.27Q, Planning for Science Activities. Any deviations from the technical work plan are documented in the following sections as they occur. The TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model assessments: (1) Mechanical and thermal impacts of basalt magma intrusion on the invert, waste packages and waste forms of the intersected emplacement drifts of Zone 1. (2) Temperature and pressure trends of basaltic magma intrusion intersecting Zone 1 and their potential effects on waste packages and waste forms in Zone 2 emplacement drifts. (3) Deleterious volatile gases, exsolving from the intruded basalt magma and their potential effects on waste packages of Zone 2 emplacement drifts. (4) Post-intrusive physical

  16. Hydrogen generation in tru waste transportation packages

    SciTech Connect

    Anderson, B; Sheaffer, M K; Fischer, L E

    2000-03-27

    This document addresses hydrogen generation in TRU waste transportation packages. The potential sources of hydrogen generation are summarized with a special emphasis on radiolysis. After defining various TRU wastes according to groupings of material types, bounding radiolytic G-values are established for each waste type. Analytical methodologies are developed for prediction of hydrogen gas concentrations for various packaging configurations in which hydrogen generation is due to radiolysis. Representative examples are presented to illustrate how analytical procedures can be used to estimate the hydrogen concentration as a function of time. Methodologies and examples are also provided to show how the time to reach a flammable hydrogen concentration in the innermost confinement layer can be estimated. Finally, general guidelines for limiting the hydrogen generation in the payload and hydrogen accumulation in the innermost confinement layer are described.

  17. Packaged low-level waste verification system

    SciTech Connect

    Tuite, K.T.; Winberg, M.; Flores, A.Y.; Killian, E.W.; McIsaac, C.V.

    1996-08-01

    Currently, states and low-level radioactive waste (LLW) disposal site operators have no method of independently verifying the radionuclide content of packaged LLW that arrive at disposal sites for disposal. At this time, disposal sites rely on LLW generator shipping manifests and accompanying records to insure that LLW received meets the waste acceptance criteria. An independent verification system would provide a method of checking generator LLW characterization methods and help ensure that LLW disposed of at disposal facilities meets requirements. The Mobile Low-Level Waste Verification System (MLLWVS) provides the equipment, software, and methods to enable the independent verification of LLW shipping records to insure that disposal site waste acceptance criteria are being met. The MLLWVS system was developed under a cost share subcontract between WMG, Inc., and Lockheed Martin Idaho Technologies through the Department of Energy`s National Low-Level Waste Management Program at the Idaho National Engineering Laboratory (INEL).

  18. Drift emplaced waste package thermal response

    SciTech Connect

    Ruffner, D.J.; Johnson, G.L.; Platt, E.A.; Blink, J.A.; Doering, T.W.

    1993-01-01

    Thermal calculations of the effects of radioactive waste decay heat on the I repository at Yucca Mountain, Nevada have been conducted by the Yucca Mountain Site Characterization Project (YMP) at Lawrence Livermore National Laboratory (LLNL) in conjunction with the B&W Fuel Company. For a number of waste package spacings, these 3D transient calculations use the TOPAZ3D code to predict drift wall temperatures to 10,000 years following emplacement. Systematic tcniperature variation occurs as a function of fuel age at emplacement and Areal Mass Loading (AML) during the first few centuries after emplacement. After about 1000 years, emplacement age is not a strong driver on rock temperature; AML has a larger impact. High AMLs occur when large waste packages are emplaced end-tocnd in drifts. Drift emplacement of equivalent packages results in lower rock teniperatures than borehole emplacement. For an emplacement scheme with 50% of the drift length occupied by packages, an AML of 138 MTU/acre is about three times higher than the Site Characterization Plan-Conceptual Design (SCP-CD) value. With this higher AML (requiring only 1/3 of the SCP-CD repository footprint), peak drift wall temperatures do not exceed 160*C, but rock temperatures excetd the boiling point of water for about 3000 years. These TOPAZ3D results Iiive been compared with reasonable agreement with two other computer codes.

  19. Drift emplaced waste package thermal response

    SciTech Connect

    Ruffner, D.J.; Johnson, G.L.; Platt, E.A.; Blink, J.A.; Doering, T.W.

    1993-12-31

    Thermal calculations of the effects of radioactive waste decay heat on the potential repository at Yucca Mountain, Nevada, have been conducted by the Yucca Mountain Site Characterization Project (YMP) at Lawrence Livermore National Lab. (LLNL) in conjunction with the B&W Fuel Co. For a number of waste package spacings, these 3D transient calculations use the TOPAZ3D code to predict drift wall temperatures to 10,000 years following emplacement. Systematic temperature variation occurs as a function of fuel age at emplacement and Areal Mass Loading (AML) during the first few centuries after emplacement. After about 1000 years, emplacement age is not a strong driver on rock temperature; AML has a larger impact. High AMLs occur when large waste packages are emplaced end-to-end in drifts. Drift emplacement of equivalent packages results in lower rock temperatures than borehole emplacement. For an emplacement scheme with 50% of the drift length occupied by packages, an AML of 138 MTU/acre is about three times higher than the Site Characterization Plan-Conceptual Design (SCP-CD) value. With this higher AML (requiring only 1/3 of the SCP-CD repository footprint), peak drift wall temperatures do not exceed 160{degrees}C, but rock temperatures exceed the boiling point of water for about 3000 years. These TOPAZ3D results have been compared with reasonable agreement with two other computer codes.

  20. Packaged digester for treating animal wastes

    SciTech Connect

    Not Available

    1981-11-03

    A new range of packaged digesters to process animal or organic wastes has been developed by Bovis Civil Engineering. The unit, known as the Polygester is suitable for use on factory farms, isolated communities and manufacturing industries. The unit consists of an anaerobic digester together with associated pumps, heat exchangers and pipework ready-assembled on a rigid common chassis and separate gas holder as a packaged system. Based on an undiluted solids input of 11% pig slurry, performance figures show up to 85% reduction of COD, 95% reduction of BOD and 18 m3 of biogas per day (equivalent to about 10 litres fuel oil).

  1. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    SciTech Connect

    S. LeStrange

    1999-11-15

    The Monitored Geologic Repository Waste Package Operations of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Navy (Refs. 1 and 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package. The water may gradually leach the fissile components and neutron absorbers out of the waste package. In addition, the accumulation of silica (SiO{sub 2}) in the waste package over time may further affect the neutronics of the system. This study presents calculations of the long-term geochemical behavior of waste packages containing the Enhanced Design Alternative (EDA) II inner shell, Navy canister, and basket components. The calculations do not include the Navy SNF in the waste package. The specific study objectives were to determine the chemical composition of the water and the quantity of silicon (Si) and other solid corrosion products in the waste package during the first million years after the waste package is breached. The results of this calculation will be used to ensure that the type and amount of criticality control material used in the waste package design will prevent criticality.

  2. Mixed waste chemical compatibility with packaging components

    SciTech Connect

    Nigrey, P.J.; Conroy, M.; Blalock, L.B.

    1994-05-01

    In this paper, a chemical compatibility testing program for packaging of mixed wastes at will be described. We will discuss the choice of four y-radiation doses, four time durations, four temperatures and four waste solutions to simulate the hazardous waste components of mixed wastes for testing materials compatibility of polymers. The selected simulant wastes are (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. A selection of 10 polymers with anticipated high resistance to one or more of these types of environments are proposed for testing as potential liner or seal materials. These polymers are butadiene acrylonitrile copolymer, cross-linked polyethylene, epichlorhyarin, ethylene-propylene rubber, fluorocarbon, glass-filled tetrafluoroethylene, high-density poly-ethylene, isobutylene-isoprene copolymer, polypropylene, and styrene-butadiene rubber. We will describe the elements of the testing plan along with a metric for establishing time resistance of the packaging materials to radiation and chemicals.

  3. 44 BWR Waste Package Loading Curve Evaluation

    SciTech Connect

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  4. Industrial Waste Landfill IV upgrade package

    SciTech Connect

    Not Available

    1994-03-29

    The Y-12 Plant, K-25 Site, and ORNL are managed by DOE`s Operating Contractor (OC), Martin Marietta Energy Systems, Inc. (Energy Systems) for DOE. Operation associated with the facilities by the Operating Contractor and subcontractors, DOE contractors and the DOE Federal Building result in the generation of industrial solid wastes as well as construction/demolition wastes. Due to the waste streams mentioned, the Y-12 Industrial Waste Landfill IV (IWLF-IV) was developed for the disposal of solid industrial waste in accordance to Rule 1200-1-7, Regulations Governing Solid Waste Processing and Disposal in Tennessee. This revised operating document is a part of a request for modification to the existing Y-12 IWLF-IV to comply with revised regulation (Rule Chapters 1200-1-7-.01 through 1200-1-7-.08) in order to provide future disposal space for the ORR, Subcontractors, and the DOE Federal Building. This revised operating manual also reflects approved modifications that have been made over the years since the original landfill permit approval. The drawings referred to in this manual are included in Drawings section of the package. IWLF-IV is a Tennessee Department of Environmental and Conservation/Division of Solid Waste Management (TDEC/DSWM) Class 11 disposal unit.

  5. Packaging of radioactive wastes for sea disposal

    NASA Astrophysics Data System (ADS)

    The Convention on the Prevention of Marine Pollution by the Dumping of Wastes and Other Matter, known as the London Dumping Convention was adopted by an inter-governmental conference in London in 1972 and came into force in 1975. In 1977, the IAEA Board of Governors agreed that there is a continuing responsibility for the IAEA to contribute to the effectiveness of the London Dumping Conventions by providing guidance relevant to the various aspects of dumping radioactive wastes at sea. In the light of the above responsibilities, the IAEA organized a Technical Committee Meeting from 3 to 7 December 1979 to assess the current situation concerning the requirements and the practices of packaging radioactive wastes for dumping at sea with a view to providing further guidance on this subject. The results of this meeting are summarized.

  6. 10 CFR 60.143 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Monitoring and testing waste packages. 60.143 Section 60.143 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES Performance Confirmation Program § 60.143 Monitoring and testing waste packages....

  7. 10 CFR 63.134 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Monitoring and testing waste packages. 63.134 Section 63.134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A... testing waste packages. (a) A program must be established at the geologic repository operations area...

  8. WASTE PACKAGE CORROSION STUDIES USING SMALL MOCKUP EXPERIMENTS

    SciTech Connect

    B.E. Anderson; K.B. Helean; C.R. Bryan; P.V. Brady; R.C. Ewing

    2005-10-19

    The corrosion of spent nuclear fuel and subsequent mobilization of radionuclides is of great concern in a geologic repository, particularly if conditions are oxidizing. Corroding A516 steel may offset these transport processes within the proposed waste packages at the Yucca Mountain Repository (YMR) by retaining radionuclides, creating locally reducing conditions, and reducing porosity. Ferrous iron, Fe{sup 2+}, has been shown to reduce UO{sub 2}{sup 2+} to UO{sub 2(s)} [1], and some ferrous iron-bearing ion-exchange materials adsorb radionuclides and heavy metals [2]. Of particular interest is magnetite, a potential corrosion product that has been shown to remove TcO{sub 4}{sup -} from solution [3]. Furthermore, if Fe{sup 2+} minerals, rather than fully oxidized minerals such as goethite, are produced during corrosion, then locally reducing conditions may be present. High electron availability leads to the reduction and subsequent immobilization of problematic dissolved species such as TcO{sub 4}{sup -}, NpO{sub 2}{sup +}, and UO{sub 2}{sup 2+} and can also inhibit corrosion of spent nuclear fuel. Finally, because the molar volume of iron material increases during corrosion due to oxygen and water incorporation, pore space may be significantly reduced over long time periods. The more water is occluded, the bulkier the corrosion products, and the less porosity is available for water and radionuclide transport. The focus of this paper is on the nature of Yucca Mountain waste package steel corrosion products and their effects on local redox state, radionuclide transport, and porosity.

  9. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    J.K. Knudson

    2003-10-02

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

  10. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    A. Alsaed

    2005-07-28

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

  11. Cleanup Verification Package for the 300-8 Waste Site

    SciTech Connect

    J. M. Capron

    2005-11-07

    This cleanup verification package documents completion of remedial action for the 300-8 waste site. This waste site was formerly used to stage scrap metal from the 300 Area in support of a program to recycle aluminum.

  12. Initial waste package interaction tests: status report

    SciTech Connect

    Shade, J.W.; Bradley, D.J.

    1980-12-01

    This report describes the results of some initial investigations of the effects of rock media on the release of simulated fission products from a sngle waste form, PNL reference glass 76-68. All tests assemblies contained a minicanister prepared by pouring molten, U-doped 76-68 glass into a 2-cm-dia stanless steel tube closed at one end. The tubes were cut to 2.5 to 7.5 cm in length to expose a flat glass surface rimmed by the canister wall. A cylindrical, whole rock pellet, cut from one of the rock materials used, was placed on the glass surface then both the canister and rock pellet were packed in the same type of rock media ground to about 75 ..mu..m to complete the package. Rock materials used were a quartz monzonite basalt and bedded salt. These packages were run from 4 to 6 weeks in either 125 ml digestion bombs or 850 ml autoclaves capable of direct solution sampling, at either 250 or 150/sup 0/C. Digestion bomb pressures were the vapor pressure of water, 600 psig at 250/sup 0/C, and the autoclaves were pressurized at 2000 psig with an argon overpressure. In general, the solution chemistry of these initial package tests suggests that the rock media is the dominant controlling factor and that rock-water interaction may be similar to that observed in some geothermal areas. In no case was uranium observed in solution above 15 ppB. The observed leach rates of U glass not in contact with potential sinks (rock surfaces and alteration products) have been observed to be considerably higher. Thus the use of leach rates and U concentrations observed from binary leach experiments (waste-form water only) to ascertain long-term environmental consequences appear to be quite conservative compared to actual U release in the waste package experiments. Further evaluation, however, of fission product transport behavior and the role of alteration phases as fission product sinks is required.

  13. Depleted uranium as a backfill for nuclear fuel waste package

    DOEpatents

    Forsberg, Charles W.

    1998-01-01

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  14. Depleted uranium as a backfill for nuclear fuel waste package

    DOEpatents

    Forsberg, C.W.

    1998-11-03

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  15. Depleted uranium as a backfill for nuclear fuel waste package

    SciTech Connect

    Forsberg, Charles W.

    1997-12-01

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotonically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  16. Waste forms, packages, and seals working group summary

    SciTech Connect

    Sridhar, N.; McNeil, M.B.

    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.

  17. Waste package/repository impact study: Final report

    SciTech Connect

    Not Available

    1985-09-01

    The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs.

  18. Parametric thermal evaluations of waste package emplacement

    SciTech Connect

    Bahney, R.H. III; Doering, T.W.

    1996-02-01

    Parametric thermal evaluations of spent nuclear fuel (SNF) waste packages (WPs) emplaced in the potential repository were performed to determine the impact of thermal loading, WP spacing, drift diameter, SNF aging, backfill, and relocation on the design of the Engineered Barrier System. Temperatures in the WP and near-field host rock are key to radionuclide containment, as they directly affect oxidation rates of the metal barriers and the ability of the rock to impede particle movement which must be demonstrated for a safe and licensable repository. Maximum allowable temperatures are based on material performance criteria and are specified as the following design goals for the WP/EBS design: SNF cladding 350{degrees}C, drift wall 200{degrees}C, and TSw3 rock 115{degrees}C.

  19. Remote Handling Equipment for a High-Level Waste Waste Package Closure System

    SciTech Connect

    Kevin M. Croft; Scott M. Allen; Mark W. Borland

    2006-04-01

    High-level waste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL.

  20. Packaging and transportation manual. Chapter on the packaging and transportation of hazardous and radioactive waste

    SciTech Connect

    1998-03-01

    The purpose of this chapter is to outline the requirements that Los Alamos National Laboratory employees and contractors must follow when they package and ship hazardous and radioactive waste. This chapter is applied to on-site, intra-Laboratory, and off-site transportation of hazardous and radioactive waste. The chapter contains sections on definitions, responsibilities, written procedures, authorized packaging, quality assurance, documentation for waste shipments, loading and tiedown of waste shipments, on-site routing, packaging and transportation assessment and oversight program, nonconformance reporting, training of personnel, emergency response information, and incident and occurrence reporting. Appendices provide additional detail, references, and guidance on packaging for hazardous and radioactive waste, and guidance for the on-site transport of these wastes.

  1. Cleanup Verification Package for the 600-259 Waste Site

    SciTech Connect

    J. M. Capron

    2006-02-09

    This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste.

  2. General Corrosion and Localized Corrosion of Waste Package Outer Barrier

    SciTech Connect

    K.G. Mon

    2004-10-01

    The waste package design for the License Application is a double-wall waste package underneath a protective drip shield (BSC 2004 [DIRS 168489]; BSC 2004 [DIRS 169480]). The purpose and scope of this model report is to document models for general and localized corrosion of the waste package outer barrier (WPOB) to be used in evaluating waste package performance. The WPOB is constructed of Alloy 22 (UNS N06022), a highly corrosion-resistant nickel-based alloy. The inner vessel of the waste package is constructed of Stainless Steel Type 316 (UNS S31600). Before it fails, the Alloy 22 WPOB protects the Stainless Steel Type 316 inner vessel from exposure to the external environment and any significant degradation. The Stainless Steel Type 316 inner vessel provides structural stability to the thinner Alloy 22 WPOB. Although the waste package inner vessel would also provide some performance for waste containment and potentially decrease the rate of radionuclide transport after WPOB breach before it fails, the potential performance of the inner vessel is far less than that of the more corrosion-resistant Alloy 22 WPOB. For this reason, the corrosion performance of the waste package inner vessel is conservatively ignored in this report and the total system performance assessment for the license application (TSPA-LA). Treatment of seismic and igneous events and their consequences on waste package outer barrier performance are not specifically discussed in this report, although the general and localized corrosion models developed in this report are suitable for use in these scenarios. The localized corrosion processes considered in this report are pitting corrosion and crevice corrosion. Stress corrosion cracking is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]).

  3. Production of iron from metallurgical waste

    SciTech Connect

    Hendrickson, David W; Iwasaki, Iwao

    2013-09-17

    A method of recovering metallic iron from iron-bearing metallurgical waste in steelmaking comprising steps of providing an iron-bearing metallurgical waste containing more than 55% by weight FeO and FeO equivalent and a particle size of at least 80% less than 10 mesh, mixing the iron-bearing metallurgical waste with a carbonaceous material to form a reducible mixture where the carbonaceous material is between 80 and 110% of the stoichiometric amount needed to reduce the iron-bearing waste to metallic iron, and as needed additions to provide a silica content between 0.8 and 8% by weight and a ratio of CaO/SiO.sub.2 between 1.4 and 1.8, forming agglomerates of the reducible mixture over a hearth material layer to protect the hearth, heating the agglomerates to a higher temperature above the melting point of iron to form nodules of metallic iron and slag material from the agglomerates by melting.

  4. Nuclear waste package design for the Vadose zone in tuff

    SciTech Connect

    O`Neal, W.C.; Ballou, L.B.; Gregg, D.W.; Russell, E.W.

    1984-02-01

    This report presents an overview of the selection and analysis of conceptual waste package designs that will be used by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for disposal of high-level nuclear waste (HLW) at the proposed Yucca Mountain, Nevada Site. The design requirements that the waste packages are required to meet are listed. Concept drawings for the reference designs and one alternative package design are shown. Four metal alloys; 304L SS, 321 SS, 316L SS and Incoloy 825 have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and economic analysis supporting the selection of the conceptual waste package designs is included. Post-closure containment and release rates are not discussed in this paper. 17 references, 2 figures, 2 tables.

  5. Rethink Disposable: Packaging Waste Source Reduction Pilot Project

    EPA Pesticide Factsheets

    Information about the SFBWQP Rethink Disposable: Packaging Waste Source Reduction Pilot Project, part of an EPA competitive grant program to improve SF Bay water quality focused on restoring impaired waters and enhancing aquatic resources.

  6. Cleanup Verification Package for the 300-18 Waste Site

    SciTech Connect

    J. M. Capron

    2005-08-26

    This cleanup verification package documents completion of remedial action for the 300-18 waste site. This site was identified as containing radiologically contaminated soil, metal shavings, nuts, bolts, and concrete.

  7. REMOTE MATERIAL HANDLING IN THE YUCCA MOUNTAIN WASTE PACKAGE CLOSURE CELL AND SUPPORT AREA GLOVEBOX

    SciTech Connect

    K.M. Croft; S.M. Allen; M.W. Borland

    2005-08-02

    The Yucca Mountain Waste Package Closure System (WPCS) cells provide for shielding of highly radioactive materials contained in unsealed waste packages. The purpose of the cells is to provide safe environments for package handling and sealing operations. Once sealed, the packages are placed in the Yucca Mountain Repository. Closure of a typical waste package involves a number of remote operations. Those involved typically include the placement of matched lids onto the waste package. The lids are then individually sealed to the waste package by welding. Currently, the waste package includes three lids. One lid is placed before movement of the waste package to the closure cell; the final two are placed inside the closure cell, where they are welded to the waste package. These and other important operations require considerable remote material handling within the cell environment. This paper discusses the remote material handling equipment, designs, functions, operations, and maintenance, relative to waste package closure.

  8. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    SciTech Connect

    T.L. Mitchell

    2000-05-31

    The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS M&O 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS M&O 2000a).

  9. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect

    Washington TRU Solutions LLC

    2003-03-21

    This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure also provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal.

  10. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect

    Washington TRU Solutions LLC

    2003-08-28

    This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure also provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal.

  11. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect

    Washington TRU Solutions LLC

    2003-05-06

    This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure also provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal.

  12. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect

    Washington TRU Solutions LLC

    2002-12-18

    This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure also provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal.

  13. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect

    Washington TRU Solutions LLC

    2002-10-17

    This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal.

  14. Computerized waste-accountability shipping and packaging system. [WASP

    SciTech Connect

    Jackson, J.A.; Baston, M. Jr.; DeVer, E.A.

    1981-01-01

    The Waste Accountability, Shipping and Packaging System (WASP) is a real-time computerized system designed and implemented by Mound Facility to meet the stringent packaging and reporting requirements of radioactive waste being shipped to burial sites. The system stores packaging data and inspection results for each unit and prepares all necessary documents at the time of shipment. Shipping data specific for each burial site are automatically prepared on magnetic tape for transmission to the computing center at that site. WASP has enabled Mound Facility to effectively meet the requirements of the burial sites, diminishing the possibility of being rejected from a site because of noncompliance.

  15. Iron phosphate compositions for containment of hazardous metal waste

    DOEpatents

    Day, D.E.

    1998-05-12

    An improved iron phosphate waste form for the vitrification, containment and long-term disposition of hazardous metal waste such as radioactive nuclear waste is provided. The waste form comprises a rigid iron phosphate matrix resulting from the cooling of a melt formed by heating a batch mixture comprising the metal waste and a matrix-forming component. The waste form comprises from about 30 to about 70 weight percent P{sub 2}O{sub 5} and from about 25 to about 50 weight percent iron oxide and has metals present in the metal waste chemically dissolved therein. The concentration of iron oxide in the waste form along with a high proportion of the iron in the waste form being present as Fe{sup 3+} provide a waste form exhibiting improved chemical resistance to corrosive attack. A method for preparing the improved iron phosphate waste forms is also provided. 21 figs.

  16. Iron phosphate compositions for containment of hazardous metal waste

    DOEpatents

    Day, Delbert E.

    1998-01-01

    An improved iron phosphate waste form for the vitrification, containment and long-term disposition of hazardous metal waste such as radioactive nuclear waste is provided. The waste form comprises a rigid iron phosphate matrix resulting from the cooling of a melt formed by heating a batch mixture comprising the metal waste and a matrix-forming component. The waste form comprises from about 30 to about 70 weight percent P.sub.2 O.sub.5 and from about 25 to about 50 weight percent iron oxide and has metals present in the metal waste chemically dissolved therein. The concentration of iron oxide in the waste form along with a high proportion of the iron in the waste form being present as Fe.sup.3+ provide a waste form exhibiting improved chemical resistance to corrosive attack. A method for preparing the improved iron phosphate waste forms is also provided.

  17. 10 CFR 63.134 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Monitoring and testing waste packages. 63.134 Section 63.134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A GEOLOGIC REPOSITORY AT YUCCA MOUNTAIN, NEVADA Performance Confirmation Program § 63.134 Monitoring...

  18. 10 CFR 63.134 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Monitoring and testing waste packages. 63.134 Section 63.134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A GEOLOGIC REPOSITORY AT YUCCA MOUNTAIN, NEVADA Performance Confirmation Program § 63.134 Monitoring...

  19. 10 CFR 63.134 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Monitoring and testing waste packages. 63.134 Section 63.134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A GEOLOGIC REPOSITORY AT YUCCA MOUNTAIN, NEVADA Performance Confirmation Program § 63.134 Monitoring...

  20. 10 CFR 63.134 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Monitoring and testing waste packages. 63.134 Section 63.134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A GEOLOGIC REPOSITORY AT YUCCA MOUNTAIN, NEVADA Performance Confirmation Program § 63.134 Monitoring...

  1. Measurement of radionuclides in waste packages

    SciTech Connect

    Brodzinski, R.L.; Perkins, R.W.; Rieck, H.G.; Wogman, N.A.

    1986-10-14

    A method is described for non-destructively assaying the radionuclide content of solid waste in a sealed container comprising: (a) determining the type of reactor responsible for the generation of the waste; (b) determining the degree of fuel enrichment of the reactor; (c) determining the decay time of the waste since exposure in the reactor; (d) measuring the gamma-ray spectrum of the waste; (e) determining the matrix density of the waste from ratios of photopeak pairs from individual radioisotopes in the gamma-ray spectrum; (f) measuring the total neutron flux and the coincidence neutron flux of the waste; and (g) determining neutron multiplicity from the coincidence neutron flux.

  2. Conceptual waste packaging options for deep borehole disposal

    SciTech Connect

    Su, Jiann -Cherng; Hardin, Ernest L.

    2015-07-01

    This report presents four concepts for packaging of radioactive waste for disposal in deep boreholes. Two of these are reference-size packages (11 inch outer diameter) and two are smaller (5 inch) for disposal of Cs/Sr capsules. All four have an assumed length of approximately 18.5 feet, which allows the internal length of the waste volume to be 16.4 feet. However, package length and volume can be scaled by changing the length of the middle, tubular section. The materials proposed for use are low-alloy steels, commonly used in the oil-and-gas industry. Threaded connections between packages, and internal threads used to seal the waste cavity, are common oilfield types. Two types of fill ports are proposed: flask-type and internal-flush. All four package design concepts would withstand hydrostatic pressure of 9,600 psi, with factor safety 2.0. The combined loading condition includes axial tension and compression from the weight of a string or stack of packages in the disposal borehole, either during lower and emplacement of a string, or after stacking of multiple packages emplaced singly. Combined loading also includes bending that may occur during emplacement, particularly for a string of packages threaded together. Flask-type packages would be fabricated and heat-treated, if necessary, before loading waste. The fill port would be narrower than the waste cavity inner diameter, so the flask type is suitable for directly loading bulk granular waste, or loading slim waste canisters (e.g., containing Cs/Sr capsules) that fit through the port. The fill port would be sealed with a tapered, threaded plug, with a welded cover plate (welded after loading). Threaded connections between packages and between packages and a drill string, would be standard drill pipe threads. The internal flush packaging concepts would use semi-flush oilfield tubing, which is internally flush but has a slight external upset at the joints. This type of tubing can be obtained with premium, low

  3. Humid air corrosion of YMP waste package candidate material

    SciTech Connect

    Gdowski, G.E.

    1998-01-01

    The Yucca Mountain Site Characterization Project is evaluating candidate materials for high level nuclear waste containers (Waste Packages) for a potential deep geologic repository at Yucca Mountain, Nevada. The potential repository is located above the water table in the unsaturated zone. The rock contains nominally 10% by volume water and gas pressure in the emplacement drifts of the repository is expected to remain near the ambient atmospheric pressure. The heat generated by the radioactive decay of the waste will raise the temperature of the waste packages and the surrounding rock. Waste Package temperatures above the ambient boiling point of water are anticipated for the waste emplacement scenarios. Because the repository emplacement drifts are expected to remain at the ambient atmospheric pressure, the maximum relative humidity obtainable decreases above the boiling point of water. Temperatures of the Waste Packages and the surrounding rock are expected to reach maximum temperature within 100`s of years and then gradually decrease with time. Episodic liquid water contact with the WPs is also expected; this will result in the deposition of salts and mineral scale.

  4. Recovery and distribution of incinerated aluminum packaging waste.

    PubMed

    Hu, Y; Bakker, M C M; de Heij, P G

    2011-12-01

    A study was performed into relations between physical properties of aluminum packaging waste and the corresponding aluminum scraps in bottom ash from three typical incineration processes. First, Dutch municipal solid waste incineration (MSWI) bottom ash was analyzed for the identifiable beverage can alloy scraps in the +2mm size ranges using chemical detection and X-ray fluorescence. Second, laboratory-scale pot furnace tests were conducted to investigate the relations between aluminum packaging in base household waste and the corresponding metal recovery rates. The representative packaging wastes include beverage cans, foil containers and thin foils. Third, small samples of aluminum packaging waste were incinerated in a high-temperature oven to determine leading factors influencing metal recovery rates. Packaging properties, combustion conditions, presence of magnesium and some specific contaminants commonly found in household waste were investigated independently in the high-temperature oven. In 2007, the bottom ash (+2mm fraction) from the AEB MSWI plant was estimated to be enriched by 0.1 wt.% of aluminum beverage cans scrap. Extrapolating from this number, the recovery potential of all eleven MSWI plants in the Netherlands is estimated at 720 ton of aluminum cans scrap. More than 85 wt.% of this estimate would end up in +6mm size fractions and were amenable for efficient recycling. The pot furnace tests showed that the average recovery rate of metallic aluminum typically decreases from beverage cans (93 wt.%) to foil containers (85 wt.%) to thin foils (77 wt.%). The oven tests showed that in order of decreasing impact the main factors promoting metallic aluminum losses are the packaging type, combustion temperature, residence time and salt contamination. To a lesser degree magnesium as alloying element, smaller packaging size and basic contaminations may also promote losses.

  5. Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

    SciTech Connect

    Soo, P.

    1984-08-01

    The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables.

  6. Review of DOE waste package program. Subtask 1.1 - National Waste Package Program, October 1983-March 1984. Volume 6

    SciTech Connect

    Soo, P.

    1985-03-01

    The present effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluation of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, tuff, and granite repositories. In the current Biannual Report a review of progress in the new crystalline repository (granite) program is described. Other foreign data for this host rock have also been outlined where relevant. The use of crushed salt, and bentonite- and zeolite-containing packing materials is discussed. The effects of temperature and gamma irradiation are shown to be important with respect to defining the localized environmental conditions around a waste package and the long-term integrity of the packing.

  7. Peer Review of the Waste Package Material Performance Interim Report

    SciTech Connect

    J. A. Beavers; T. M. Devine, Jr.; G. S. Frankel; R. H. Jones; R. G. Kelly; R. M. Latanision; J. H. Payer

    2001-09-04

    At the request of the U.S. Department of Energy, Bechtel SAIC Company, LLC, formed the Waste Package Materials Performance Peer Review Panel (the Panel) to review the technical basis for evaluating the long-term performance of waste package materials in a proposed repository at Yucca Mountain, Nevada. This is the interim report of the Panel; a final report will be issued in February 2002. In its work to date, the Panel has identified important issues regarding waste package materials performance. In the remainder of its work, the Panel will address approaches and plans to resolve these issues. In its review to date, the Panel has not found a technical basis to conclude that the waste package materials are unsuitable for long-term containment at the proposed Yucca Mountain Repository. Nevertheless, significant technical issues remain unsettled and, primarily because of the extremely long life required for the waste packages, there will always be some uncertainty in the assessment. A significant base of scientific and engineering knowledge for assessing materials performance does exist and, therefore, the likelihood is great that uncertainty about the long-term performance can be substantially reduced through further experiments and analysis.

  8. A history of solid waste packaging at the Hanford Site

    SciTech Connect

    Duncan, D.R.; Weyns-Rollosson, D.I.; Pottmeyer, J.A.; Stratton, T.J.

    1995-02-01

    Since the initiation of the defense materials product mission, a total of more than 600,000 m{sup 3} of radioactive solid waste has been stored or disposed at the US Department of Energy`s (DOE) Hanford Site, located in southeastern Washington State. As the DOE complex prepares for its increasing role in environmental restoration and waste remediation, the characterization of buried and retrievably stored waste will become increasingly important. Key to this characterization is an understanding of the standards and specifications to which waste was packaged; the regulations that mandated these standards and specifications; the practices used for handling and packaging different waste types; and the changes in these practices with time.

  9. Effects of mixed waste simulants on transportation packaging plastic components

    SciTech Connect

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

    1994-12-31

    The purpose of hazardous and radioactive materials packaging is to, enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified. The design requirements for both hazardous and radioactive materials packaging specify packaging compatibility, i.e., that the materials of the packaging and any contents be chemically compatible with each other. Furthermore, Type A and Type B packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program, supported by the US Department of Energy`s (DOE) Transportation Management Division, EM-261 provides the means to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, we describe the general elements of the testing program and the experimental results of the screening tests. The implications of the results of this testing are discussed in the general context of packaging development. Additionally, we present the results of the first phase of this experimental program. This phase involved the screening of five candidate liner and six seal materials against four simulant mixed wastes.

  10. Measurement of radionuclides in waste packages

    DOEpatents

    Brodzinski, Ronald L.; Perkins, Richard W.; Rieck, Henry G.; Wogman, Ned A.

    1986-01-01

    A method is described for non-destructively assaying the radionuclide content of solid waste in a sealed container by analysis of the waste's gamma-ray spectrum and neutron emissions. Some radionuclides are measured by characteristic photopeaks in the gamma-ray spectrum; transuranic nuclides are measured by neutron emission rate; other radionuclides are measured by correlation with those already measured.

  11. Measurement of radionuclides in waste packages

    DOEpatents

    Brodzinski, R.L.; Perkins, R.W.; Rieck, H.G.; Wogman, N.A.

    1984-09-12

    A method is described for non-destructively assaying the radionuclide content of solid waste in a sealed container by analysis of the waste's gamma-ray spectrum and neutron emissions. Some radionuclides are measured by characteristic photopeaks in the gamma-ray spectrum; transuranic nuclides are measured by neutron emission rate; other radionuclides are measured by correlation with those already measured.

  12. WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT

    SciTech Connect

    M. C. Knapp

    1999-09-23

    The waste package (WP) closure weld development task is part of a larger engineering development program to develop waste package designs. The purpose of the larger waste package engineering development program is to develop nuclear waste package fabrication and closure methods that the Nuclear Regulatory Commission will find acceptable and will license for disposal of spent nuclear fuel (SNF), non-fuel components, and vitrified high-level waste within a Monitored Geologic Repository (MGR). Within the WP closure development program are several major development tasks, which, in turn, are divided into subtasks. The major tasks include: WP fabrication development, WP closure weld development, nondestructive examination (NDE) development, and remote in-service inspection development. The purpose of this report is to present the objectives, technical information, and work scope relating to the WP closure weld development.and NDE tasks and subtasks and to report results of the closure weld and NDE development programs for fiscal year 1999 (FY-99). The objective of the FY-99 WP closure weld development task was to develop requirements for closure weld surface and volumetric NDE performance demonstrations, investigate alternative NDE inspection techniques, and develop specifications for welding, NDE, and handling system integration. In addition, objectives included fabricating several flat plate mock-ups that could be used for NDE development, stress relief peening, corrosion testing, and residual stress testing.

  13. Strategy for experimental validation of waste package performance assessment

    SciTech Connect

    Bates, J.K.; Abrajano, T.A. Jr.; Wronkiewicz, D.J.; Gerding, T.J.; Seils, C.A.

    1990-07-01

    A strategy for the experimental validation of waste package performance assessment has been developed as part of a program supported by the Repository Technology Program. The strategy was developed by reviewing the results of laboratory analog experiments, in-situ tests, repository simulation tests, and material interaction tests. As a result of the review, a listing of dependent and independent variables that influence the ingress of water into the near-field environment, the reaction between water and the waste form, and the transport of radionuclides from the near-field environment was developed. The variables necessary to incorporate into an experimental validation strategy were chosen by identifying those which had the greatest effect of each of the three major events, i.e., groundwater ingress, waste package reactions, and radionuclide transport. The methodology to perform validation experiments was examined by utilizing an existing laboratory analog approach developed for unsaturated testing of glass waste forms. 185 refs., 9 figs., 2 tabs.

  14. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages

    SciTech Connect

    Bacon, Diana H.; Pierce, Eric M.; Wellman, Dawn M.; Strachan, Denis M.; Josephson, Gary B.

    2006-07-31

    The primary purpose of the work reported here is to analyze the potential effect of the release of technetium (Tc) from metal inclusions in bulk vitrification waste packages once they are placed in the Integrated Disposal Facility (IDF). As part of the strategy for immobilizing waste from the underground tanks at Hanford, selected wastes will be immobilized using bulk vitrification. During analyses of the glass produced in engineering-scale tests, metal inclusions were found in the glass product. This report contains the results from experiments designed to quantify the corrosion rates of metal inclusions found in the glass product from AMEC Test ES-32B and simulations designed to compare the rate of Tc release from the metal inclusions to the release of Tc from glass produced with the bulk vitrification process. In the simulations, the Tc in the metal inclusions was assumed to be released congruently during metal corrosion as soluble TcO4-. The experimental results and modeling calculations show that the metal corrosion rate will, under all conceivable conditions at the IDF, be dominated by the presence of the passivating layer and corrosion products on the metal particles. As a result, the release of Tc from the metal particles at the surfaces of fractures in the glass releases at a rate similar to the Tc present as a soluble salt. The release of the remaining Tc in the metal is controlled by the dissolution of the glass matrix. To summarize, the release of 99Tc from the BV glass within precipitated Fe is directly proportional to the diameter of the Fe particles and to the amount of precipitated Fe. However, the main contribution to the Tc release from the iron particles is over the same time period as the release of the soluble Tc salt. For the base case used in this study (0.48 mass% of 0.5 mm diameter metal particles homogeneously distributed in the BV glass), the release of 99Tc from the metal is approximately the same as the release from 0.3 mass% soluble Tc

  15. Secondary Waste Form Down Selection Data Package – Ceramicrete

    SciTech Connect

    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 Laboratory 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

  16. WAPDEG Analysis of Waste Package and Drip shield Degradation

    SciTech Connect

    K. Mon

    2004-09-29

    As directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), an analysis of the degradation of the engineered barrier system (EBS) drip shields and waste packages at the Yucca Mountain repository is developed. The purpose of this activity is to provide the TSPA with inputs and methodologies used to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. This analysis provides information useful to satisfy ''Yucca Mountain Review Plan, Final Report'' (NRC 2003 [DIRS 163274]) requirements. Several features, events, and processes (FEPs) are also discussed (Section 6.2, Table 15). The previous revision of this report was prepared as a model report in accordance with AP-SIII.10Q, Models. Due to changes in the role of this report since the site recommendation, it no longer contains model development. This revision is prepared as a scientific analysis in accordance with AP-SIII.9Q, ''Scientific Analyses'' and uses models previously validated in (1) ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]); (2) ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' (BSC 2004 [DIRS 169984]); and (3) ''General Corrosion and Localized Corrosion of Drip Shield'' (BSC 2004 [DIRS 169845]). The integrated waste package degradation (IWPD) analysis presented in this report treats several implementation-related issues, such as defining the number and size of patches per waste package that undergo stress corrosion cracking; recasting the weld flaw analysis in a form as implemented in the Closure Weld Defects (CWD) software; and, general corrosion rate manipulations (e.g., change of scale in Section 6.3.4). The weld flaw portion of this report takes input from an engineering calculation (BSC 2004

  17. Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package

    SciTech Connect

    S.F.A. Deng; M. Saglam; L.J. Gratton

    2001-05-23

    In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

  18. Cleanup Verification Package for the 300 VTS Waste Site

    SciTech Connect

    S. W. Clark and T. H. Mitchell

    2006-03-13

    This cleanup verification package documents completion of remedial action for the 300 Area Vitrification Test Site, also known as the 300 VTS site. The site was used by Pacific Northwest National Laboratory as a field demonstration site for in situ vitrification of soils containing simulated waste.

  19. Cleanup Verification Package for the 600-47 Waste Site

    SciTech Connect

    M. J. Cutlip

    2005-08-26

    This cleanup verification package documents completion of interim remedial action for the 600-47 waste site. This site consisted of several areas of surface debris and contamination near the banks of the Columbia River across from Johnson Island. Contaminated material identified in field surveys included four areas of soil, wood, nuts, bolts, and other metal debris.

  20. 10 CFR 60.143 - Monitoring and testing waste packages.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... GEOLOGIC REPOSITORIES Performance Confirmation Program § 60.143 Monitoring and testing waste packages. (a) A program shall be established at the geologic repository operations area for monitoring the... be emplaced in the underground facility. (b) Consistent with safe operation at the...

  1. INITIAL WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: UNCANISTERED FUEL (TBV)

    SciTech Connect

    J.R. Massari

    1995-10-06

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide an assessment of the present waste package design from a criticality risk standpoint, The specific objectives of this initial analysis are to: (1) Establish a process for determining the probability of waste package criticality as a function of time (in terms of a cumulative distribution function, probability distribution function, or expected number of criticalities in a specified time interval) for various waste package concepts; (2) Demonstrate the established process by estimating the probability of criticality as a function of time since emplacement for an intact uncanistered fuel waste package (UCF-WP) configuration; and (3) Identify the dominant sequences leading to waste package criticality for subsequent detailed analysis. The purpose of this analysis is to document and demonstrate the developed process as it has been applied to the UCF-WP. This revision is performed to correct deficiencies in the previous revision and provide further detail on the calculations performed. Due to the current lack of knowledge in a number of areas, every attempt has been made to ensure that the all calculations and assumptions were conservative. This analysis is preliminary in nature, and is intended to be superseded by at least two more versions prior to license application. The information and assumptions used to generate this analysis are unverified and have been globally assigned TBV identifier TBV-059-WPD. Future versions of this analysis will update these results, possibly replacing the global TBV with a small number of TBV's on individual items, with the goal of removing all TBV designations by license application submittal. The final output of this document, the probability of UCF-WP criticality as a function of time, is therefore, also TBV. This document is intended to deal only with the risk of internal criticality with unaltered fuel

  2. Hydrophilic Packaging of Iron Oxide Nanoclusters for Highly Sensitive Imaging

    PubMed Central

    Smith, Cartney E.; Ernenwein, Dawn; Shkumatov, Artem; Clay, Nicholas; Lee, JuYeon; Melhem, Molly; Misra, Sanjay; Zimmerman, Steven C.; Kong, Hyunjoon

    2015-01-01

    Superparamagnetic iron oxide nanoparticles (SPIONs) are used as imaging probes to provide contrast in magnetic resonance images. Successful use of SPIONs in targeted applications greatly depends on their ability to generate contrast, even at low levels of accumulation, in the tissue of interest. In the present study, we report that SPION nanoclusters packaged to a controlled size by a hyperbranched polyglycerol (HPG) can target tissue defects and have a high relaxivity of 719 mM−1s−1, which was close to their theoretical maximal limit. The resulting nanoclusters were able to identify regions of defective vasculature in an ischemic murine hindlimb using MRI with iron doses that were 5–10 fold lower than those typically used in preclinical studies. Such high relaxivity was attributed to the molecular architecture of HPG, which mimics that of the water retentive polysaccharide, glycogen. The results of this study will be broadly useful in sensitive imaging applications. PMID:26291408

  3. Preclosure analysis of conceptual waste package designs for a nuclear waste repository in tuff

    SciTech Connect

    O`Neal, W.C.; Gregg, D.W.; Hockman, J.N.; Russell, E.W.; Stein, W.

    1984-11-01

    This report discusses the selection and analysis of conceptual waste package developed by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for possible disposal of high-level nuclear waste at a candidate site at Yucca Mountain, Nevada. The design requirements that the waste package must conform to are listed, as are several desirable design considerations. Illustrations of the reference and alternative designs are shown. Four austenitic stainless steels (316L SS, 321 SS, 304L SS and Incoloy 825 high nickel alloy) have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and ecnonmic analyses supporting the selection of the conceptual waste package designs is included. Postclosure containment and release rates are not analyzed in this report.

  4. Oxidation and waste-to-energy output of aluminium waste packaging during incineration: A laboratory study.

    PubMed

    López, Félix A; Román, Carlos Pérez; García-Díaz, Irene; Alguacil, Francisco J

    2015-09-01

    This work reports the oxidation behaviour and waste-to-energy output of different semi-rigid and flexible aluminium packagings when incinerated at 850°C in an air atmosphere enriched with 6% oxygen, in the laboratory setting. The physical properties of the different packagings were determined, including their metallic aluminium contents. The ash contents of their combustion products were determined according to standard BS ISO 1171:2010. The net calorific value, the required energy, and the calorific gain associated with each packaging type were determined following standard BS EN 13431:2004. Packagings with an aluminium lamina thickness of >50μm did not fully oxidise. During incineration, the weight-for-weight waste-to-energy output of the packagings with thick aluminium lamina was lower than that of packagings with thin lamina. The calorific gain depended on the degree of oxidation of the metallic aluminium, but was greater than zero for all the packagings studied. Waste aluminium may therefore be said to act as an energy source in municipal solid waste incineration systems.

  5. Number of Waste Package Hit by Igneous Intrusion

    SciTech Connect

    M. Wallace

    2004-10-13

    The purpose of this scientific analysis report is to document calculations of the number of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analyses include disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application (TSPA-LA) for the Yucca Mountain Project (YMP). Igneous activity is a disruptive event that is included in the TSPA-LA analyses. Two igneous activity scenarios are considered: (1) The igneous intrusion groundwater release scenario (also called the igneous intrusion scenario) considers the in situ damage to waste packages or failure of waste packages that occurs if they are engulfed or otherwise affected by magma as a result of an igneous intrusion. (2) The volcanic eruption scenario depicts the direct release of radioactive waste due to an intrusion that intersects the repository followed by a volcanic eruption at the surface. An igneous intrusion is defined as the ascent of a basaltic dike or dike system (i.e., a set or swarm of multiple dikes comprising a single intrusive event) to repository level, where it intersects drifts. Magma that does reach the surface from igneous activity is an eruption (or extrusive activity) (Jackson 1997 [DIRS 109119], pp. 224, 333). The objective of this analysis is to develop a probabilistic measure of the number of waste packages that could be affected by each of the two scenarios.

  6. SPENT NUCLEAR FUEL WASTE PACKAGE FILLER TESTING TECHNICAL GUIDELINES DOCUMENT

    SciTech Connect

    J.A. Cogar

    1996-08-28

    The purpose of this work is to provide supporting calculations for determination of the radiation source terms specific to subsurface shielding design and analysis. These calculations are not intended to provide the absolute values of the source terms, which are under the charter of the Waste Package Operations (WPO) Group. Rather, the calculations focus on evaluation of the various combinations of fuel enrichment, burnup and cooling time for a given decay heat output, consistent with the waste package (WP) thermal design basis. The objective is to determine the worst-case combination of the fuel characteristics (enrichment, burnup and cooling time) which would give the maximum radiation fields for subsurface shielding considerations. The calculations are limited to PWR fuel only, since the WP design is currently evolving with thinner walls and a reduced heat load as compared to the viability assessment (VA) reference design. The results for PWR fuel will provide a comparable indication of the trend for BWR fuel, as their characteristics are similar. The source term development for defense high-level waste and other spent nuclear fuel (SNF) is the responsibility of the WPO Group, and therefore, is not included this work. This work includes the following items responsive to the stated purpose and objective: (1) Determine the possible fuel parameters (initial enrichment, burnup and cooling time), that give the same decay heat value as specified for the waste package thermal design; (2) Obtain the neutron and gamma source terms for the various combinations of the fuel parameters for use in radiation field calculations; and (3) Calculate radiation fields on the surfaces of the waste package and its transporter to quantify the effects of the fuel parameters with the same decay heat value for use in identifying the worst-case combination of the fuel parameters.

  7. Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

    SciTech Connect

    Manaktala, H.K. . Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. . Div. of High-Level Waste Management)

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

  8. Microbial Effects on Nuclear Waste Packaging Materials

    SciTech Connect

    Horn, J; Martin, S; Carrillo, C; Lian, T

    2005-07-22

    Microorganisms may enhance corrosion of components of planned engineered barriers within the proposed nuclear waste repository at Yucca Mountain (YM). Corrosion could occur either directly, through processes collectively known as Microbiologically Influenced Corrosion (MIC), or indirectly, by adversely affecting the composition of water or brines that come into direct contact with engineered barrier surfaces. Microorganisms of potential concern (bacteria, archea, and fungi) include both those indigenous to Yucca Mountain and those that infiltrate during repository construction and after waste emplacement. Specific aims of the experimental program to evaluate the potential of microorganisms to affect damage to engineered barrier materials include the following: Indirect Effects--(1) Determine the limiting factors to microbial growth and activity presently in the YM environment. (2) Assess these limiting factors to aid in determining the conditions and time during repository evolution when MIC might become operant. (3) Evaluate present bacterial densities, the composition of the YM microbial community, and determining bacterial densities if limiting factors are overcome. During a major portion of the regulatory period, environmental conditions that are presently extant become reestablished. Therefore, these studies ascertain whether biomass is sufficient to cause MIC during this period and provide a baseline for determining the types of bacterial activities that may be expected. (4) Assess biogenic environmental effects, including pH, alterations to nitrate concentration in groundwater, the generation of organic acids, and metal dissolution. These factors have been shown to be those most relevant to corrosion of engineered barriers. Direct Effects--(1) Characterize and quantify microbiological effects on candidate containment materials. These studies were carried out in a number of different approaches, using whole YM microbiological communities, a subset of YM

  9. How Waste Management Can Be Influenced By Transport Packagings

    SciTech Connect

    Roland, V.

    2002-02-28

    With major D&D projects ongoing or being planned, and also with the daily management of radwaste from nuclear facilities, the potential role of transport packagings has often been overlooked: here will one rely essentially on drums, there several local waste processing units are built, elsewhere decommissioned facilities are cut in small bits to fit into small containers by far less efficient. The present paper proposes to illustrate how integrating a transport system from the start may influence operational choices of waste management.

  10. Secondary Waste Cementitious Waste Form Data Package for the Integrated Disposal Facility Performance Assessment

    SciTech Connect

    Cantrell, Kirk J.; Westsik, Joseph H.; Serne, R Jeffrey; Um, Wooyong; Cozzi, Alex D.

    2016-05-16

    A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at the Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.

  11. Compatibility of packaging components with simulant mixed waste

    SciTech Connect

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

    1996-04-01

    The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the US have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (US DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). Based on these national requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program provides a basis to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, the authors present the results of the second phase of this testing program. The first phase screened five liner materials and six seal materials towards four simulant mixed wastes. This phase involved the comprehensive testing of five candidate liner materials to an aqueous Hanford Tank simulant mixed waste. The comprehensive testing protocol involved exposing the respective materials a matrix of four gamma radiation doses ({approximately} 1, 3, 6, and 40 kGy), three temperatures (18, 50, and 60 C), and four exposure times (7, 14, 28, and 180 days). Following their exposure to these combinations of conditions, the materials were evaluated by measuring five material properties. These properties were specific gravity, dimensional changes, hardness, stress cracking, and mechanical properties.

  12. EXTERNAL CRITICALITY CALCULATION FOR DOE SNF CODISPOSAL WASTE PACKAGES

    SciTech Connect

    H. Radulescu

    2002-10-18

    The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St. Vrain, Melt and Dilute, Shippingport Light Water Breeder Reactor (LWBR), N-Reactor, and Training, Research, Isotope, General Atomics reactor (TRIGA). The results of this calculation are intended to be used for estimating the probability of criticality in the near-field and in the far-field. There are no limitations on use of the results of this calculation. The calculation is associated with the waste package design and was developed in accordance with the technical work plan, ''Technical Work Plan for: Department of Energy Spent Nuclear Fuel and Plutonium Disposition Work Packages'' (Bechtel SAIC Company, LLC [BSC], 2002a). This calculation is subject to the Quality Assurance Requirements and Description (QARD) per the activity evaluation under work package number P6212310Ml in the technical work plan TWP-MGR-MD-0000 10 REV 01 (BSC 2002a).

  13. Aging and Phase Stability of Waste Package Outer Barrier

    SciTech Connect

    Tammy S. Edgecumble Summers

    2001-08-23

    This Analysis Model Report (AMR) was prepared in accordance with the Work Direction and Planning Document, ''Aging and Phase Stability of Waste Package Outer Barrier'' (CRWMS M&O 1999a). ICN 01 of this AMR was developed following guidelines provided in TWP-MGR-MD-000004 REV 01, ''Technical Work Plan for: Integrated Management of Technical Product Input Department'' (BSC 2001, Addendum B). It takes into consideration the Enhanced Design Alternative II (EDA II), which has been selected as the preferred design for the Engineered Barrier System (EBS) by the License Application Design Selection (LADS) program team (CRWMS M&O 1999b). The salient features of the EDA II design for this model are a waste package (WP) consisting of an outer barrier of Alloy 22 and an inner barrier of Type 316L stainless steel. This report provides information on the phase stability of Alloy 22l, the current waste-package-outer-barrier (WPOB) material. These phase stability studies are currently divided into three general areas: (1) Long-range order reactions; (2) Intermetallic and carbide precipitation in the base metal; and (3) Intermetallic and carbide precipitation in welded samples.

  14. Materials of Criticality Safety Concern in Waste Packages

    SciTech Connect

    Larson, S.L.; Day, B.A.

    2006-07-01

    10 CFR 71.55 requires in part that the fissile material package remain subcritical when considering 'the most reactive credible configuration consistent with the chemical and physical form of the material'. As waste drums and packages may contain unlimited types of materials, determination of the appropriately bounding moderator and reflector materials to ensure compliance with 71.55 requires a comprehensive analysis. Such an analysis was performed to determine the materials or elements that produce the most reactive configuration with regards to both moderation and reflection of a Pu-239 system. The study was originally performed for the TRUPACT-II shipping package and thus the historical fissile mass limit for the package, 325 g Pu-239, was used [1]. Reactivity calculations were performed with the SCALE package to numerically assess the moderation or reflection merits of the materials [2]. Additional details and results are given in SAIC-1322-001 [3]. The development of payload controls utilizing process knowledge to determine the classification of special moderator and/or reflector materials and the associated fissile mass limit is also addressed. (authors)

  15. Development of integrated radioactive waste packaging and conditioning solutions in the UK

    SciTech Connect

    Sibley, Peter; Butter, Kevin; Zimmerman, Ian; Viermann, Joerg; Messer, Matthias

    2013-07-01

    In order to offer a more cost effective, safer and efficient Intermediate Level Waste (ILW) management service, EnergySolutions EU Ltd. and Gesellschaft fur Nuklear-Service mbH (GNS) have been engaged in the development of integrated radioactive waste retrieval, packaging and conditioning solutions in the UK. Recognising the challenges surrounding regulatory endorsement and on-site implementation in particular, this has resulted in an alternative approach to meeting customer, safety regulator and disposability requirements. By working closely with waste producers and the organisation(s) responsible for endorsing radioactive waste management operations in the UK, our proposed solutions are now being implemented. By combining GNS' off-the-shelf, proven Ductile Cast Iron Containers (DCICs) and water removal technologies, with EnergySolutions EU Ltd.'s experience and expertise in waste retrieval, safety case development and disposability submissions, a fully integrated service offering has been developed. This has involved significant effort to overcome technical challenges such as onsite equipment deployment, active commissioning, conditioning success criteria and disposability acceptance. Our experience in developing such integrated solutions has highlighted the importance of working in collaboration with all parties to achieve a successful and viable outcome. Ultimately, the goal is to ensure reliable, safe and effective delivery of waste management solutions. (authors)

  16. Performance of Nonmigratory Iron Chelating Active Packaging Materials in Viscous Model Food Systems.

    PubMed

    Roman, Maxine J; Decker, Eric A; Goddard, Julie M

    2015-09-01

    Many packaged food products undergo quality deterioration due to iron promoted oxidative reactions. Recently, we have developed a nonmigratory iron chelating active packaging material that represents a novel approach to inhibit oxidation of foods while addressing consumer demands for "cleanˮ labels. A challenge to the field of nonmigratory active packaging is ensuring that surface-immobilized active agents retain activity in a true food system despite diffusional limitations. Yet, the relationship between food viscosity and nonmigratory active packaging activity retention has never been characterized. The objective of this study was to investigate the influence of food viscosity on iron chelation by a nonmigratory iron chelating active packaging material. Methyl cellulose was added to aqueous buffered iron solutions to yield model systems with viscosities ranging from ∼1 to ∼10(5)  mPa·s, representing viscosities ranging from beverage to mayonnaise. Iron chelation was quantified by material-bound iron content using colorimetry and inductively coupled plasma-optical emission spectrometry (ICP-OES).  Maximum iron chelation was reached in solutions up to viscosity ∼10(2)  mPa·s. In more viscous solutions (up to ∼10(4)  mPa·s), there was a significant decrease in iron chelating capacity (P < 0.05). However, materials still retained at least 76% iron chelating capacity. Additionally, the influence of different food hydrocolloids on the performance of nonmigratory iron chelating active packaging was characterized. Methyl cellulose and carrageenan did not compete with the material for specific iron chelation (P > 0.05). Materials retained 32% to 45% chelating capacity when in contact with competitively chelating hydrocolloids guar gum, locust bean gum, and xanthan gum. This work demonstrates the potential application of nonmigratory iron chelating active packaging in liquid and semi-liquid foods to allow for the removal of synthetic chelators, while

  17. The Role of Packaging in Solid Waste Management 1966 to 1976.

    ERIC Educational Resources Information Center

    Darnay, Arsen; Franklin, William E.

    The goals of waste processors and packagers obviously differ: the packaging industry seeks durable container material that will be unimpaired by external factors. Until recently, no systematic analysis of the relationship between packaging and solid waste disposal had been undertaken. This three-part document defines these interactions, and the…

  18. 10 CFR 60.135 - Criteria for the waste package and its components.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for the waste package and its components. 60.135 Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135...

  19. 10 CFR 60.135 - Criteria for the waste package and its components.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for the waste package and its components. 60.135 Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135...

  20. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    SciTech Connect

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  1. EVALUATION OF WASTE PACKAGE EXTERNAL ENVIRONMENTAL CONDITION STUDY

    SciTech Connect

    E. N. Lindner and E. F. Dembowski

    1998-07-23

    The U. S. Department of Energy (DOE) is studying Yucca Mountain as the possible site for a permanent underground repository for disposal of spent nuclear fuel (SNF) and other high-level waste (HLW). The emplacement of high-level radioactive waste in Yucca Mountain will release a large amount of heat into the rock above and below the repository. Due to this heat, the rock temperature will rise, and then decrease when the production of decay heat falls below the rate at which heat escapes from the hot zone. In addition to raising the rock temperature, the heat will vaporize water, which will condense in cooler regions. The condensate water may drain back toward the emplacement drifts or it may ''shed'' through the pillars between emplacement drifts. Other effects, such as coupled chemical and mechanical processes, may influence the movement of water above, within, and below the emplacement drifts. This study examined near field environmental parameters that could have an effect on the waste package, including temperature, humidity, seepage rate, pH of seepage, chemistry (dissolved salts/minerals) of seepage, composition of drift atmosphere, colloids, and biota. This report is a Type I analysis performed in support of the development of System Description Documents (SDDs). A Type I analysis is a quantitative or qualitative analysis that may fulfill any of a variety of purposes associated with the Monitored Geologic Repository (MGR), other than providing direct analytical support for design output documents. A Type I analysis may establish design input, as defined in the ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998). This study establishes a technical basis for emplacement drift (i.e. at the waste package surface) environment criteria to be considered in the development of the waste package design. The information will support development of several SDDs and resolve emplacement drift external environment questions in the criteria of those

  2. ANALYSIS OF DAMAGE TO WASTE PACKAGES CAUSED BY SEISMIC EVENTS DURING POST-CLOSURE

    SciTech Connect

    Alves, S W; Blair, S C; Carlson, S R; Gerhard, M; Buscheck, T A

    2008-05-27

    This paper presents methodology and results of an analysis of damage due to seismic ground motion for waste packages emplaced in a nuclear waste repository at Yucca Mountain, Nevada. A series of three-dimensional rigid body kinematic simulations of waste packages, pallets, and drip shields subjected to seismic ground motions was performed. The simulations included strings of several waste packages and were used to characterize the number, location, and velocity of impacts that occur during seismic ground motion. Impacts were categorized as either waste package-to-waste package (WP-WP) or waste package-to-pallet (WP-P). In addition, a series of simulations was performed for WP-WP and WP-P impacts using a detailed representation of a single waste package. The detailed simulations were used to determine the amount of damage from individual impacts, and to form a damage catalog, indexed according to the type, angle, location and force/velocity of the impact. Finally, the results from the two analyses were combined to estimate the total damage to a waste package that may occur during an episode of seismic ground motion. This study addressed two waste package types, four levels of peak ground velocity (PGV), and 17 ground motions at each PGV. Selected aspects of waste package degradation, such as effective wall thickness and condition of the internals, were also considered. As expected, increasing the PGV level of the vibratory ground motion increases the damage to the waste packages. Results show that most of the damage is caused by WP-P impacts. TAD-bearing waste packages with intact internals are highly resistant to damage, even at a PGV of 4.07 m/s, which is the highest level analyzed.

  3. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages. Erratum

    SciTech Connect

    Smith, Gary L.

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  4. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    SciTech Connect

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  5. Iron Phosphate Glasses for Vitrifying DOE High Priority Nuclear Wastes

    SciTech Connect

    Kim, C.W.; Day, D.E.

    2004-03-29

    Iron phosphate glasses have been studied as an alternative glass for vitrifying Department of Energy (DOE) high priority wastes. The high priority wastes were the Low Activity Waste (LAW) and the High Level Waste (HLW) with high chrome content stored at Hanford, WA, and the Sodium Bearing Waste (SBW) stored at the Idaho National Engineering and Environmental Laboratory. These wastes were recommended by Tanks Focus Area since they were expected to require special attention when vitrified in borosilicate glasses. All three of these wastes have been successfully vitrified in iron phosphate glasses at waste loadings ranging from a low of 32 wt% for the high sulfate LAW to 40 wt% for the SBW to a high of 75 wt% for the high chrome HLW. In addition to these desirable high waste loadings, the iron phosphate glasses were easily melted, typically between 950 and 1200 C, in less than 4 hours in commercial refractory oxide containers. It is noteworthy that the chemical durability of both glassy and deliberately crystallized iron phosphate wasteforms not only met, but significantly exceeded, all current DOE chemical durability requirements as measured by the Product Consistency Test (PCT) and Vapor Hydration Test (VHT). The high waste loading, low melting temperature, rapid furnace throughput (short melting time) and their outstanding chemical durability could significantly accelerate the clean up effort and reduce the time and cost of vitrifying these high priority wastes.

  6. Production patterns of packaging waste categories generated at typical Mediterranean residential building worksites.

    PubMed

    González Pericot, N; Villoria Sáez, P; Del Río Merino, M; Liébana Carrasco, O

    2014-11-01

    The construction sector is responsible for around 28% of the total waste volume generated in Europe, which exceeds the amount of household waste. This has led to an increase of different research studies focusing on construction waste quantification. However, within the research studies made, packaging waste has been analyzed to a limited extent. This article focuses on the packaging waste stream generated in the construction sector. To this purpose current on-site waste packaging management has been assessed by monitoring ten Mediterranean residential building works. The findings of the experimental data collection revealed that the incentive measures implemented by the construction company to improve on-site waste sorting failed to achieve the intended purpose, showing low segregation ratios. Subsequently, through an analytical study the generation patterns for packaging waste are established, leading to the identification of the prevailing kinds of packaging and the products responsible for their generation. Results indicate that plastic waste generation maintains a constant trend throughout the whole construction process, while cardboard becomes predominant towards the end of the construction works with switches and sockets from the electricity stage. Understanding the production patterns of packaging waste will be beneficial for adapting waste management strategies to the identified patterns for the specific nature of packaging waste within the context of construction worksites.

  7. Analysis of Ecodesign Implementation and Solutions for Packaging Waste System by Using System Dynamics Modeling

    NASA Astrophysics Data System (ADS)

    Berzina, Alise; Dace, Elina; Bazbauers, Gatis

    2010-01-01

    This paper discusses the findings of a research project which explored the packaging waste management system in Latvia. The paper focuses on identifying how the policy mechanisms can promote ecodesign implementation and material efficiency improvement and therefore reduce the rate of packaging waste accumulation in landfill. The method used for analyzing the packaging waste management policies is system dynamics modeling. The main conclusion is that the existing legislative instruments can be used to create an effective policy for ecodesign implementation but substantially higher tax rates on packaging materials and waste disposal than the existing have to be applied.

  8. Assessment of actinide mass embedded in large concrete waste packages by photon interrogation and photofission.

    PubMed

    Gmar, M; Jeanneau, F; Lainé, F; Makil, H; Poumarède, B; Tola, F

    2005-01-01

    This paper describes a method based on photofission developed in our laboratory to characterize in depth large waste packages. The method consists in using photons of high-energy (Bremsstrahlung radiation) in order to induce reactions of photofission on the heavy nuclei present in the wastes. The measurement of the delayed neutrons allows quantifying the actinides in the wastes. We present the first results of measurement performed with a concrete mock-up of 870l and two real waste packages.

  9. Types of packaging waste from secondary sources (supermarkets)--the situation in the UK.

    PubMed

    Dixon-Hardy, Darron W; Curran, Beverley A

    2009-03-01

    Packaging waste is a contributing factor to the large quantity of waste that is sent to landfill in the UK. This research focuses on waste from the secondary packaging sector in the UK. In particular, supermarkets were investigated as they supply a large section of consumers with their grocery and other requirements and generate high quantities of packaging waste due to the high turnover within the store. In general, supermarkets use either metal cages or wooden pallets to transport products from depot to store. Investigation shows that packaging waste produced when using the wooden pallets is greater than for metal cages but the use of wooden pallets allows for greater versatility when in the store. The type of transit packaging used depends on what the products are initially packaged in and how the supermarket supply chain works. All cardboard and high-grade plastic is recycled but, depending on the facilities at the stores, the low-grade plastic can be recycled as well. This paper details types of packaging used within the supermarket secondary packaging sector and how waste can be reduced. To reduce the amount of packaging waste produced by the supermarkets, the products will have to be wrapped differently by the producers so that less packaging is needed in transit.

  10. Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

    SciTech Connect

    Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

    1982-02-01

    In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

  11. Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

    SciTech Connect

    M. J. Appel and J. M. Capron

    2007-07-25

    This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

  12. Addendum to the Safety Analysis Report for the Steel Waste Packaging. Revision 1

    SciTech Connect

    Crow, S R

    1996-02-15

    The Battelle Pacific Northwest National Laboratory Safety Analysis Report (SAR) for the Steel Waste Package requires additional analyses to support the shipment of remote-handled radioactive waste and special-case waste from the 324 building hot cells to PUREX for interim storage. This addendum provides the analyses required to show that this waste can be safely shipped onsite in the configuration shown.

  13. Production patterns of packaging waste categories generated at typical Mediterranean residential building worksites

    SciTech Connect

    González Pericot, N.; Villoria Sáez, P.; Del Río Merino, M.; Liébana Carrasco, O.

    2014-11-15

    Highlights: • On-site segregation level: 1.80%; training and motivation strategies were not effective. • 70% Cardboard waste: from switches and sockets during the building services stage. • 40% Plastic waste: generated during structures and partition works due to palletizing. • >50% Wood packaging waste, basically pallets, generated during the envelope works. - Abstract: The construction sector is responsible for around 28% of the total waste volume generated in Europe, which exceeds the amount of household waste. This has led to an increase of different research studies focusing on construction waste quantification. However, within the research studies made, packaging waste has been analyzed to a limited extent. This article focuses on the packaging waste stream generated in the construction sector. To this purpose current on-site waste packaging management has been assessed by monitoring ten Mediterranean residential building works. The findings of the experimental data collection revealed that the incentive measures implemented by the construction company to improve on-site waste sorting failed to achieve the intended purpose, showing low segregation ratios. Subsequently, through an analytical study the generation patterns for packaging waste are established, leading to the identification of the prevailing kinds of packaging and the products responsible for their generation. Results indicate that plastic waste generation maintains a constant trend throughout the whole construction process, while cardboard becomes predominant towards the end of the construction works with switches and sockets from the electricity stage. Understanding the production patterns of packaging waste will be beneficial for adapting waste management strategies to the identified patterns for the specific nature of packaging waste within the context of construction worksites.

  14. Cleanup Verification Package for the 118-B-1, 105-B Solid Waste Burial Ground

    SciTech Connect

    J. M. Capron

    2008-01-21

    This cleanup verification package documents completion of remedial action, sampling activities, and compliance criteria for the 118-B-1, 105-B Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-B Reactor and P-10 Tritium Separation Project and also received waste from the 105-N Reactor. The burial ground received reactor hardware, process piping and tubing, fuel spacers, glassware, electrical components, tritium process wastes, soft wastes and other miscellaneous debris.

  15. Reduction experiment of iron scale by adding waste plastics.

    PubMed

    Zhang, Chongmin; Chen, Shuwen; Miao, Xincheng; Yuan, Hao

    2009-01-01

    The special features of waste plastics in China are huge in total amount, various in type and dispersive in deposition. Therefore, it is necessary to try some new ways that are fit to Chinese situation for disposing waste plastics as metallurgical raw materials more effectively and flexibly. Owing to its high ferrous content and less impurity, the iron scale became ideal raw material to produce pure iron powder. One of the methods to produce pure iron powder is Hoganas Method, by which, after one or multistage of reduction steps, the iron scale can be reduced pure iron powder. However, combining utilization of waste plastics and iron powder production, a series of reduction experiments were arranged and investigated, which is hoped to take use of both thermal and chemical energy contained in waste plastics as well as to improve the reducing condition of iron scale, and hence to develop a new metallurgical way of disposing waste plastics. The results show that under these experimental conditions, the thermal-decomposition of water plastics can conduce to an increase of porosity in the reduction systems. Moreover, better thermodynamics and kinetics conditions for the reduction of scale can be reached. As a result, the reduction rate is increased.

  16. Depleted uranium oxides as spent-nuclear-fuel waste-package fill materials

    SciTech Connect

    Forsberg, C.W.

    1997-07-07

    Depleted uranium dioxide fill inside the waste package creates the potential for significant improvements in package performance based on uranium geochemistry, reduces the potential for criticality in a repository, and consumes DU inventory. As a new concept, significant uncertainties exist: fill properties, impacts on package design, post- closure performance.

  17. Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment

    SciTech Connect

    McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O'Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

    2001-02-01

    This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

  18. Thermal Evaluation for the Naval SNF Waste Package

    SciTech Connect

    T.L. Mitchell

    2000-04-25

    The purpose of this calculation is to evaluate the thermal performance of the naval long spent nuclear fuel (SNF) waste package (WP) under multiple disposal conditions in a monitored geologic repository (MGR). The scope of this calculation is limited to determination of thermal temperature profiles upon the surface of, and within, the naval long SNF WP. The objective is to develop a temperature profile history within the WP, at time increments up to 10,000 years of emplacement. The results of this calculation are intended to support the Naval SNF WP Analysis and Model Report (AMR) for Site Recommendation (SR). This calculation was performed to the specifications within its Technical Development Plan (TDP) (Ref. 8.16). This calculation is developed and documented in accordance with the AP-3.12Q/REV. 0IICN. 0 procedure, Calculations.

  19. UCF WASTE PACKAGE SHIELDING ANALYSIS/2-D DORT (SCPB: N/A)

    SciTech Connect

    D.J. Skulina

    1996-01-18

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the dose rates from the UCF waste packages to be used by the EBS and other repository systems to incorporate ALARA practices in the overall repository design in compliance with the goals of the Waste Package Implementation Plan for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives. The objectives of this evaluation are (1) to show the dose rate as a function of distance from the waste package surface and (2) to provide the shielding thicknesses required for the waste package transporter to meet a 10 mr/hr target dose rate at 2 meters from the transporter surface.

  20. Value Engineering Study for Closing Waste Packages Containing TAD Canisters

    SciTech Connect

    Colleen Shelton-Davis

    2005-11-01

    The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system.

  1. Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5

    SciTech Connect

    MANN, F.M.

    2000-03-02

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

  2. WASTE CONTAINER AND WASTE PACKAGE PERFORMANCE MODELING TO SUPPORT SAFETY ASSESSMENT OF LOW AND INTERMEDIATE-LEVEL RADIOACTIVE WASTE DISPOSAL.

    SciTech Connect

    SULLIVAN, T.

    2004-06-30

    Prior to subsurface burial of low- and intermediate-level radioactive wastes, a demonstration that disposal of the wastes can be accomplished while protecting the health and safety of the general population is required. The long-time frames over which public safety must be insured necessitates that this demonstration relies, in part, on computer simulations of events and processes that will occur in the future. This demonstration, known as a Safety Assessment, requires understanding the performance of the disposal facility, waste containers, waste forms, and contaminant transport to locations accessible to humans. The objective of the coordinated research program is to examine the state-of-the-art in testing and evaluation short-lived low- and intermediate-level waste packages (container and waste form) in near surface repository conditions. The link between data collection and long-term predictions is modeling. The objective of this study is to review state-of-the-art modeling approaches for waste package performance. This is accomplished by reviewing the fundamental concepts behind safety assessment and demonstrating how waste package models can be used to support safety assessment. Safety assessment for low- and intermediate-level wastes is a complicated process involving assumptions about the appropriate conceptual model to use and the data required to support these models. Typically due to the lack of long-term data and the uncertainties from lack of understanding and natural variability, the models used in safety assessment are simplistic. However, even though the models are simplistic, waste container and waste form performance are often central to the case for making a safety assessment. An overview of waste container and waste form performance and typical models used in a safety assessment is supplied. As illustrative examples of the role of waste container and waste package performance, three sample test cases are provided. An example of the impacts of

  3. Determination of Radioisotope Content by Measurement of Waste Package Dose Rates - 13394

    SciTech Connect

    Souza, Daiane Cristini B.; Gimenes Tessaro, Ana Paula; Vicente, Roberto

    2013-07-01

    The objective of this communication is to report the observed correlation between the calculated air kerma rates produced by radioactive waste drums containing untreated ion-exchange resin and activated charcoal slurries with the measured radiation field of each package. Air kerma rates at different distances from the drum surface were calculated with the activity concentrations previously determined by gamma spectrometry of waste samples and the estimated mass, volume and geometry of solid and liquid phases of each waste package. The water content of each waste drum varies widely between different packages. Results will allow determining the total activity of wastes and are intended to complete the previous steps taken to characterize the radioisotope content of wastes packages. (authors)

  4. Using Aspen simulation package to determine solubility of mixed salts in TRU waste evaporator bottoms

    SciTech Connect

    Hatchell, J.L.

    1998-03-01

    Nitric acid from plutonium process waste is a candidate for waste minimization by recycling. Process simulation software packages, such as Aspen, are valuable tools to estimate how effective recovery processes can be, however, constants in equations of state for many ionic components are not in their data libraries. One option is to combine single salt solubility`s in the Aspen model for mixed salt system. Single salt solubilities were regressed in Aspen within 0.82 weight percent of literature values. These were combined into a single Aspen model and used in the mixed salt studies. A simulated nitric acid waste containing mixed aluminum, calcium, iron, magnesium and sodium nitrate was tested to determine points of solubility between 25 and 100 C. Only four of the modeled experimental conditions, at 50 C and 75 C, produced a saturated solution. While experimental results indicate that sodium nitrate is the first salt to crystallize out, the Aspen computer model shows that the most insoluble salt, magnesium nitrate, the first salt to crystallize. Possible double salt formation is actually taking place under experimental conditions, which is not captured by the Aspen model.

  5. Aging and Phase Stability of Waste Package Outer Barrier

    SciTech Connect

    F. Wong

    2004-09-28

    This report was prepared in accordance with ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). This report provides information on the phase stability of Alloy 22, the current waste package outer barrier material. The goal of this model is to determine whether the single-phase solid solution is stable under repository conditions and, if not, how fast other phases may precipitate. The aging and phase stability model, which is based on fundamental thermodynamic and kinetic concepts and principles, will be used to provide predictive insight into the long-term metallurgical stability of Alloy 22 under relevant repository conditions. The results of this model are used by ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' as reference-only information. These phase stability studies are currently divided into three general areas: Tetrahedrally close-packed (TCP) phase and carbide precipitation in the base metal; TCP and carbide precipitation in welded samples; and Long-range ordering reactions. TCP-phase and carbide precipitates that form in Alloy 22 are generally rich in chromium (Cr) and/or molybdenum (Mo) (Raghavan et al. 1984 [DIRS 154707]). Because these elements are responsible for the high corrosion resistance of Alloy 22, precipitation of TCP phases and carbides, especially at grain boundaries, can lead to an increased susceptibility to localized corrosion in the alloy. These phases are brittle and also tend to embrittle the alloy (Summers et al. 1999 [DIRS 146915]). They are known to form in Alloy 22 at temperatures greater than approximately 600 C. Whether these phases also form at the lower temperatures expected in the repository during the 10,000-year regulatory period must be determined. The kinetics of this precipitation will be determined for both the base metal and the weld heat-affected zone (HAZ). The TCP phases (P, {mu}, and {sigma}) are

  6. Options for reducing food waste by 'Quality Controlled Logistics' using intelligent packaging along the supply chain.

    PubMed

    Heising, Jenneke K; Claassen, G D H; Dekker, Matthijs

    2017-04-04

    Optimizing supply chain management can help to reduce food waste. This article describes how intelligent packaging can be used to reduce food waste when used in supply chain management based on Quality Controlled Logistics (QCL). Intelligent packaging senses compounds in the package that correlate with the critical quality attribute of a food product. The information on the quality of each individual packaged food item that is provided by the intelligent packaging can be used for QCL. In a conceptual approach is explained that monitoring food quality by intelligent packaging sensors makes it possible to obtain information about the variation in the quality of foods and use a dynamic expiration date (IP-DED) on a food package. The conceptual approach is supported by quantitative data from simulations on the effect of using the information of intelligent packaging in supply chain management with the goal to reduce food waste. This simulation shows that QCL by using the information on the quality of products that is provided by intelligent packaging can substantially reduce food waste. When QCL is combined with dynamic pricing based on the predicted expiry dates a further waste reduction is envisaged.

  7. Evaluation of the Corrosivity of Dust Deposited on Waste Packages at Yucca Mountain, Nevada

    SciTech Connect

    C. Bryan; R. Jarek; T. Wolery; D. Shields; M. Sutton; E. Hardin; D. Barr

    2005-03-18

    Small amounts of dust will be deposited on the surfaces of waste packages in drifts at Yucca Mountain during the operational and the preclosure ventilation periods. Salts present in the dust will deliquesce as the waste packages cool and relative humidity in the drifts increases. In this paper, we evaluate the potential for brines formed by dust deliquescence to initiate and sustain localized corrosion that results in failure of the waste package outer barrier and early failure of the waste package. These arguments have been used to show that dust deliquescence-induced localized or crevice corrosion of the waste package outer barrier (Alloy 22) is of low consequence with respect to repository performance. Measured atmospheric and underground dust compositions are the basis of thermodynamic modeling and experimental studies to evaluate the likelihood of brine formation and persistence, the volume of brines that may form, and the relative corrosivity of the initial deliquescent brines and of brines modified by processes on the waste package surface. In addition, we evaluate several mechanisms that could inhibit or stifle localized corrosion should it initiate. The dust compositions considered include both tunnel dust samples from Yucca Mountain, National Airfall Deposition Program rainout data, and collected windblown dust samples. Also considered is sublimation of ammonium salts, a process that could affect dust composition prior to deliquescence. Ammonium chlorides, nitrates, and even sulfates sublimate readily into ammonia and acid gases, and will be lost from the surface of the waste package prior to deliquescence.

  8. Safety evaluation for packaging transportation of equipment for tank 241-C-106 waste sluicing system

    SciTech Connect

    Calmus, D.B.

    1994-08-25

    A Waste Sluicing System (WSS) is scheduled for installation in nd waste storage tank 241-C-106 (106-C). The WSS will transfer high rating sludge from single shell tank 106-C to double shell waste tank 241-AY-102 (102-AY). Prior to installation of the WSS, a heel pump and a transfer pump will be removed from tank 106-C and an agitator pump will be removed from tank 102-AY. Special flexible receivers will be used to contain the pumps during removal from the tanks. After equipment removal, the flexible receivers will be placed in separate containers (packagings). The packaging and contents (packages) will be transferred from the Tank Farms to the Central Waste Complex (CWC) for interim storage and then to T Plant for evaluation and processing for final disposition. Two sizes of packagings will be provided for transferring the equipment from the Tank Farms to the interim storage facility. The packagings will be designated as the WSSP-1 and WSSP-2 packagings throughout the remainder of this Safety Evaluation for Packaging (SEP). The WSSP-1 packagings will transport the heel and transfer pumps from 106-C and the WSSP-2 packaging will transport the agitator pump from 102-AY. The WSSP-1 and WSSP-2 packagings are similar except for the length.

  9. Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

    SciTech Connect

    D.S. Kimball; C.E. Sanders

    2006-02-07

    The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k{sub eff}) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions.

  10. Packaging waste prevention activities: A life cycle assessment of the effects on a regional waste management system.

    PubMed

    Nessi, Simone; Rigamonti, Lucia; Grosso, Mario

    2015-09-01

    A life cycle assessment was carried out to evaluate the effects of two packaging waste prevention activities on the overall environmental performance of the integrated municipal waste management system of Lombardia region, Italy. The activities are the use of refined tap water instead of bottled water for household consumption and the substitution of liquid detergents packaged in single-use containers by those distributed 'loose' through self-dispensing systems and refillable containers. A 2020 baseline scenario without waste prevention is compared with different waste prevention scenarios, where the two activities are either separately or contemporaneously implemented, by assuming a complete substitution of the traditional product(s). The results show that, when the prevention activities are carried out effectively, a reduction in total waste generation ranging from 0.14% to 0.66% is achieved, corresponding to a 1-4% reduction of the affected packaging waste fractions (plastics and glass). However, the improvements in the overall environmental performance of the waste management system can be far higher, especially when bottled water is substituted. In this case, a nearly 0.5% reduction of the total waste involves improvements ranging mostly between 5 and 23%. Conversely, for the substitution of single-use packaged liquid detergents (0.14% reduction of the total waste), the achieved improvements do not exceed 3% for nearly all impact categories.

  11. A history of ocean disposal of packaged low-level radioactive waste

    SciTech Connect

    Holcomb, W.F.

    1982-03-01

    Two methods are practiced throughout the world for the disposal of low-level radioactive wastes-ground burial and ocean dumping. Ocean dumping was used by the United States from 1946 to 1970; European nations have been ocean dumping since 1951, with the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development supervising the international ocean dumping operations since 1967. The European nations have dumped wastes containing over 700 000 Ci of radioactivity, whereas the United States has dumped wastes containing over 94 000 Ci. The Environmental Protection Agency (EPA) has surveyed some of the U. S. ocean dump sites and retrieved three drums of waste to assess the condition of the radioactive waste packaging. The NEA has published guidelines for packaging requirements for ocean disposal, and the EPA has a program to prepare regulations to complement the existing international and domestic broad-based regulations for packaging of radioactive wastes for ocean disposal.

  12. IRON PHOSPHATE GLASSES: AN ALTERNATIVE FOR VITRIFYING CERTAIN NUCLEAR WASTES

    SciTech Connect

    Day, Delbert E.; Kim, Cheol-Woon

    2004-06-28

    The unusual properties and beneficial characteristics of iron phosphate glasses, as viewed from the standpoint of alternative glasses for vitrifying nuclear and hazardous wastes (which contain components that make them poorly suited for vitrification in borosilicate glass), have been investigated by the University of Missouri-Rolla with support from the Environmental Management Science Program (EMSP), DOE [DEFG07- 96ER45618]. During the past year, the corrosion resistance of Inconel 690 and 693 coupons submerged in an iron phosphate melt at 1050 C for up to 155 days has been investigated to determine whether iron phosphate glasses could be melted in a Joule Heated Melter (JHM) equipped with such electrodes in the same manner as now being used to melt borosilicate glass. Substituting iron phosphate glasses for borosilicate glasses could significantly reduce the time and cost for clean up due to the higher waste loading possible in iron phosphate glass. The iron phosphate melt, which contained 30 wt% of the Hanford Low Activity Waste (LAW), did not corrode the Inconel 690 to any greater extent than what has been reported for Inconel 690 electrodes in the borosilicate melt in the JHM at the Defense Waste Processing Facility. Furthermore, Inconel 693 appeared to be an even better candidate for use in iron phosphate melts since its corrosion rate (0.7 {micro}m/day) was only about one half that (1.3 {micro}m/day) of Inconel 690. In the past year, the results of the research on iron phosphate glasses have been described in nine technical papers and one report and have been presented at four international and national meetings.

  13. Chemical and physical properties of waste package packing materials

    SciTech Connect

    Wood, M.I.; Relyea, J.F.; Lane, D.L.; Carlson, R.A.

    1983-08-01

    Data has been gathered to develop a preliminary understanding of the behavior of crushed basalt and sodium bentonite alone as well as mixtures of 75% crushed basalt-25% sodium bentonite, the current reference candidate material. The material properties investigated included: (1) chemical stability under dry thermal and hydrothermal conditions; (2) radionuclide sorption capacity and solubility limits; and (3) hydraulic conductivities as a function of material density and temperature. The primary results of these studies indicate that: (1) the phase structure and swelling potential of bentonite remain intact up to dehydration temperatures of 370{degree}C; (2) the primary hydrothermal reaction in a basalt-bentonite mixture is the alteration of basalt glass to smectites, zeolites, and quartz; (3) minor reaction of bentonite to form albite and quartz occurs with a slight enrichment of potassium in the bentonite phase; (4) the mobility of cationic radionuclides is low in the presence of basalt and bentonite under the expected reducing waste package geochemical conditions because of low solubility and high sorption; and (5) moderate density basalt-bentonite mixture ({ge}1.7 g/cm{sup 3}) are characterized by low hydraulic conductivities ({le}1 {times} 10{sup {minus}8} cm/sec).

  14. Investigation of metallic, ceramic, and polymeric materials for engineered barrier applications in nuclear-waste packages

    SciTech Connect

    Westerman, R.E.

    1980-10-01

    An effort to develop licensable engineered barrier systems for the long-term (about 1000 yr) containment of nuclear wastes under conditions of deep continental geologic disposal has been underway at Pacific Northwest Laboratory since January 1979, under the auspices of the High-Level Waste Immobilization Program. In the present work, the barrier system comprises the hard or structural elements of the package: the canister, the overpack(s), and the hole sleeve. A number of candidate metallic, ceramic, and polymeric materials were put through mechanical, corrosion, and leaching screening tests to determine their potential usefulness in barrier-system applications. Materials demonstrating adequate properties in the screening tests will be subjected to more detailed property tests, and, eventually, cost/benefit analyses, to determine their ultimate applicability to barrier-system design concepts. The following materials were investigated: two titanium alloys of Grade 2 and Grade 12; 300 and 400 series stainless steels, Inconels, Hastelloy C-276, titanium, Zircoloy, copper-nickel alloys and cast irons; total of 14 ceramic materials, including two grades of alumina, plus graphite and basalt; and polymers such as polyamide-imide, polyarylene, polyimide, polyolefin, polyphenylene sulfide, polysulfone, fluoropolymer, epoxy, furan, silicone, and ethylene-propylene terpolymer (EPDM) rubber. The most promising candidates for further study and potential use in engineered barrier systems were found to be rubber, filled polyphenylene sulfide, fluoropolymer, and furan derivatives.

  15. FABRICATION AND DEPLOYMENT OF THE 9979 TYPE AF RADIOACTIVE WASTE PACKAGING FOR THE DEPARTMENT OF ENERGY

    SciTech Connect

    Blanton, P.; Eberl, K.

    2013-10-10

    This paper summarizes the development, testing, and certification of the 9979 Type A Fissile Packaging that replaces the UN1A2 Specification Shipping Package eliminated from Department of Transportation (DOT) 49 CFR 173. The DOT Specification Package was used for many decades by the U.S. nuclear industry as a fissile waste container until its removal as an authorized container by DOT. This paper will discuss stream lining procurement of high volume radioactive material packaging manufacturing, such as the 9979, to minimize packaging production costs without sacrificing Quality Assurance. The authorized content envelope (combustible and non-combustible) as well as planned content envelope expansion will be discussed.

  16. Geochemistry Model Abstraction and Sensitivity Studies for the 21 PWR CSNF Waste Package

    SciTech Connect

    P. Bernot; S. LeStrange; E. Thomas; K. Zarrabi; S. Arthur

    2002-10-29

    The CSNF geochemistry model abstraction, as directed by the TWP (BSC 2002b), was developed to provide regression analysis of EQ6 cases to obtain abstracted values of pH (and in some cases HCO{sub 3}{sup -} concentration) for use in the Configuration Generator Model. The pH of the system is the controlling factor over U mineralization, CSNF degradation rate, and HCO{sub 3}{sup -} concentration in solution. The abstraction encompasses a large variety of combinations for the degradation rates of materials. The ''base case'' used EQ6 simulations looking at differing steel/alloy corrosion rates, drip rates, and percent fuel exposure. Other values such as the pH/HCO{sub 3}{sup -} dependent fuel corrosion rate and the corrosion rate of A516 were kept constant. Relationships were developed for pH as a function of these differing rates to be used in the calculation of total C and subsequently, the fuel rate. An additional refinement to the abstraction was the addition of abstracted pH values for cases where there was limited O{sub 2} for waste package corrosion and a flushing fluid other than J-13, which has been used in all EQ6 calculation up to this point. These abstractions also used EQ6 simulations with varying combinations of corrosion rates of materials to abstract the pH (and HCO{sub 3}{sup -} in the case of the limiting O{sub 2} cases) as a function of WP materials corrosion rates. The goodness of fit for most of the abstracted values was above an R{sup 2} of 0.9. Those below this value occurred during the time at the very beginning of WP corrosion when large variations in the system pH are observed. However, the significance of F-statistic for all the abstractions showed that the variable relationships are significant. For the abstraction, an analysis of the minerals that may form the ''sludge'' in the waste package was also presented. This analysis indicates that a number a different iron and aluminum minerals may form in the waste package other than those

  17. Yucca Mountain Waste Package Closure System Robotic Welding and Inspection System

    SciTech Connect

    C. I. Nichol; D. P. Pace; E. D. Larsen; T. R. McJunkin; D. E. Clark; M. L. Clark; K. L. Skinner; A. D. Watkins; H. B. Smartt

    2011-10-01

    The Waste Package Closure System (WPCS), for the closure of radioactive waste in canisters for permanent storage of spent nuclear fuel (SNF) and high-level waste in the Yucca Mountain Repository was designed, fabricated, and successfully demonstrated at the Idaho National Laboratory (INL). This article focuses on the robotic hardware and tools necessary to remotely weld and inspect the closure lid welds. The system was operated remotely and designed for use in a radiation field, due to the SNF contained in the waste packages being closed.

  18. Assessing microbiologically induced corrosion of waste package materials in the Yucca Mountain repository

    SciTech Connect

    Horn, J. M., LLNL

    1998-01-01

    The contribution of bacterial activities to corrosion of nuclear waste package materials must be determined to predict the adequacy of containment for a potential nuclear waste repository at Yucca Mountain (YM), NV. The program to evaluate potential microbially induced corrosion (MIC) of candidate waste container materials includes characterization of bacteria in the post-construction YM environment, determination of their required growth conditions and growth rates, quantitative assessment of the biochemical contribution to metal corrosion, and evaluation of overall MIC rates on candidate waste package materials.

  19. 77 FR 50716 - Certain Food Waste Disposers and Components and Packaging Thereof; Notice of Commission...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-22

    ... From the Federal Register Online via the Government Publishing Office INTERNATIONAL TRADE COMMISSION Certain Food Waste Disposers and Components and Packaging Thereof; Notice of Commission Determination Not to Review an Initial Determination Granting Complainant's Motions To Amend the Notice...

  20. 78 FR 1881 - Certain Food Waste Disposers and Components and Packaging Thereof; Notice of the Commission's...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-09

    ... From the Federal Register Online via the Government Publishing Office INTERNATIONAL TRADE COMMISSION Certain Food Waste Disposers and Components and Packaging Thereof; Notice of the Commission's Determination Not To Review Initial Determinations Granting Complainant's Motions To Partially Terminate...

  1. Thermal analysis of Yucca Mountain commercial high-level waste packages

    SciTech Connect

    Altenhofen, M.K.; Eslinger, P.W.

    1992-10-01

    The thermal performance of commercial high-level waste packages was evaluated on a preliminary basis for the candidate Yucca Mountain repository site. The purpose of this study is to provide an estimate for waste package component temperatures as a function of isolation time in tuff. Several recommendations are made concerning the additional information and modeling needed to evaluate the thermal performance of the Yucca Mountain repository system.

  2. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    SciTech Connect

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-05-09

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4,136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  3. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    NASA Astrophysics Data System (ADS)

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-09-01

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  4. 77 FR 23751 - Certain Food Waste Disposers and Components and Packaging Thereof; Institution of Investigation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-20

    ... COMMISSION Certain Food Waste Disposers and Components and Packaging Thereof; Institution of Investigation... components and packaging thereof by reason of (1) Infringement of the claim of U.S. Patent No. D535,850... investigation and, after the investigation, issue an exclusion order and a cease and desist order. ]...

  5. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

    SciTech Connect

    Not Available

    1983-06-01

    The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.

  6. MEASUREMENT OF BIOAVAILABLE IRON AT TWO HAZARDOUS WASTE SITES

    EPA Science Inventory

    In the past, the concentrations of iron II in monitoring wells has been used to evaluate natural attenuation processes at hazardous waste sites. Changes in the aqueous concentrations of electron acceptors/products are important to the evaluation of natural biological attenuation...

  7. Effects of illumination and packaging on non-heme iron and color attributes of sliced ham.

    PubMed

    Li, H; Li, C B; Xu, X L; Zhou, G H

    2012-08-01

    This study was designed to investigate effects of illumination and packaging on color of cooked cured sliced ham during refrigeration, and the possibility of decomposition of nitrosylheme under light and oxygen exposure. Three illumination levels and three packaging films with different oxygen transmission rates (OTRs) were used in two separate experiments during 35 days storage, and pH value, a* value, nitrosylheme, residual nitrite and non-heme iron were evaluated. Packaging OTRs had significant effects (P<0.01) on a* value, but illumination level and packaging OTR did not affect (P>0.05) nitrosylheme concentration during storage. For both groups, storage time had a significant effect (P<0.01) on a* value and nitrosylheme. Negative relationships between nitrosylheme and nitrite in the illumination group, and between nitrosylheme and non-heme iron in the packaging group were observed. Therefore, illumination level and packaging OTR had limited effects on overall pigment stability, but more discoloration and loss of redness occurred on the surface of products.

  8. Tritium Permeation Estimate from APT and CLWR-TEF Waste Packages

    SciTech Connect

    Clark, E.A.

    1999-03-18

    The amount of tritium permeating out of waste containers has been estimated for the Accelerator Production of Tritium project (APT) and for the Commercial Light Water Reactor - Tritium Extraction Facility project (CLWR-TEF). The waste packages analyzed include the Aluminum, Window, Tungsten, Lead, and Steel packages for the APT project, and the overpack of extracted Tritium Producing Burnable Absorber Rods (TPBARs) for the CLWR-TEF project. All of the tritium contained in the waste was assumed to be available as a gas in the free volume inside the waste container at the beginning of disposal, and to then permeate the stainless steel waste container. From estimates of the tritium content of each waste form, the void or free volume of the package, disposal temperature and container geometry, the amount of tritium exiting the waste container by permeation was calculated. Two tritium permeation paths were considered separately: through the entire wall surface area and through the weld area only, the weld area having reduced thickness and significantly less surface area compared to the wall area. Permeation out of the five APT waste containers at 50 degrees Celsius is mainly through the welds, and at 100 degrees Celsius is through the permeation out of the entire wall surface area. The largest maximum offgas rate from an APT waste stream at 50 degrees Celsius (estimated disposal temperature) was 1.8E-6 Ci/year from the weld of the Window waste package, and the smallest maximum offgas rate was 3.7E-5 Ci/year from the weld of the Lead waste package. Permeation from the CLWR-TEF overpack at 40 degrees Celsius is mainly through the entire wall surface area, with a maximum offgas rate of 1.3E-5 Ci/year.

  9. Technical Basis Document No. 6: Waste Package and Drip Shield Corrosion

    SciTech Connect

    Farmer, J; Pasupathi, V; Nair, P; Gordon, G; McCright, D; Gdowski, G; Carroll, S; Steinborn, T; Summers, T; Wong, F; Rebak, R; Lian, T; Ilevbare, G; Lee, J; Hua, F; Payer, J

    2003-08-01

    The waste package and drip shield will experience a wide range of interactive environmental conditions and degradation modes that will determine the overall performance of the waste package and repository. The operable modes of degradation are determined by the temperature regime of operation (region), and are summarized here. Dry-Out Region (T {ge} 120 C; 50 to 400 Years): During the pre-closure period, the waste package will be kept dry by ventilation air. During the thermal pulse, heat generated by radioactive decay will eventually increase the temperature of the waste package, drip shield and drift wall to a level above the boiling point, where the probability of seepage into drifts will become insignificant. Further heating will push the waste package surface temperature above the deliquescence point of expected salt mixtures, thereby preventing the formation of deliquescence brines from dust deposits and humid air. Phase and time-temperature-transformation diagrams predicted for Alloy 22, and validated with experimental data, indicates no significant phase instabilities (LRO and TCP precipitation) at temperatures below 300 C for 10,000 years. Neither will dry oxidation at these elevated temperatures limit waste package life. After the peak temperature is reached, the waste package will begin to cool, eventually reaching a point where deliquescence brine formation may occur. However, corrosion testing of Alloy 22 underneath such films has shown no evidence of life-limiting localized corrosion. Transition Region (120 C {ge} T {ge} 100 C; 400 to 1,000 Years): During continued cooling, the temperature of the drift wall will drop to a level close to the boiling point of the seepage brine, thus permitting the onset of seepage. Corrosion in a concentrated, possibly aggressive, liquid-phase brine, evolved through evaporative concentration, is possible while in this region. However, based upon chemical divide theory, most ({ge} 99%) of the seepage water entering the

  10. Cleanup Verification Package for the 118-B-6, 108-B Solid Waste Burial Ground

    SciTech Connect

    M. L. Proctor

    2006-06-13

    This cleanup verification package documents completion of remedial action for the 118-B-6, 108-B Solid Waste Burial Ground. The 118-B-6 site consisted of 2 concrete pipes buried vertically in the ground and capped by a concrete pad with steel lids. The site was used for the disposal of wastes from the "metal line" of the P-10 Tritium Separation Project.

  11. 10 CFR 60.135 - Criteria for the waste package and its components.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste... 10 Energy 2 2011-01-01 2011-01-01 false Criteria for the waste package and its components. 60.135 Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE...

  12. 10 CFR 60.135 - Criteria for the waste package and its components.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for the waste package and its components. 60.135 Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE...

  13. THERMAL HISTORY OF CLADDING IN A 21 PWR WASTE PACKAGE LOADED WITH AVERAGE FUEL

    SciTech Connect

    H.M. Wade

    2000-01-25

    The purpose of this calculation is to evaluate a mid-assembly axial fuel cladding temperature profile of a 21 pressurized water reactor (PWR) spent nuclear fuel (SNF) waste package (WP) loaded with average fuel assemblies and emplaced in a monitored geologic repository. This calculation is intended to evaluate Viability Assessment (VA) and Enhanced Design Alternatives (EDA) II design configurations in support of performance assessment. This calculation was developed by Waste Package Operations (WPO) under Office of Civilian Radioactive Waste Management (OCRWM) procedure AP-3.12Q, Revision 0.

  14. Life cycle assessment of a packaging waste recycling system in Portugal

    SciTech Connect

    Ferreira, S.; Cabral, M.; Cruz, N.F. da; Simões, P.; Marques, R.C.

    2014-09-15

    Highlights: • We modeled a real packaging waste recycling system. • The analysis was performed using the life cycle assessment methodology. • The 2010 situation was compared with scenarios where the materials were not recycled. • The “Baseline” scenario seems to be more beneficial to the environment. - Abstract: Life Cycle Assessment (LCA) has been used to assess the environmental impacts associated with an activity or product life cycle. It has also been applied to assess the environmental performance related to waste management activities. This study analyses the packaging waste management system of a local public authority in Portugal. The operations of selective and refuse collection, sorting, recycling, landfilling and incineration of packaging waste were considered. The packaging waste management system in operation in 2010, which we called “Baseline” scenario, was compared with two hypothetical scenarios where all the packaging waste that was selectively collected in 2010 would undergo the refuse collection system and would be sent directly to incineration (called “Incineration” scenario) or to landfill (“Landfill” scenario). Overall, the results show that the “Baseline” scenario is more environmentally sound than the hypothetical scenarios.

  15. Selection of barrier metals for a waste package in tuff

    SciTech Connect

    Russell, E.W.; McCright, R.D.; O`Neal, W.C.

    1983-09-01

    The Nevada Nuclear Waste Storage Investigation (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. LLNL is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. The selection of metal containment barriers is addressed. 13 references.

  16. Implementation of Control Measures for Radioactive Waste Packages with Respect to the Materials Composition - 12365

    SciTech Connect

    Steyer, S.; Kugel, K.; Brennecke, P.; Boetsch, W.; Gruendler, D.; Haider, C.

    2012-07-01

    In addition to the radiological characterization and control measures the materials composition has to be described and respective control measures need to be implemented. The approach to verify the materials composition depends on the status of the waste: - During conditioning of raw waste the control of the materials composition has to be taken into account. - For already conditioned waste a retrospective qualification of the process might be possible. - If retrospective process qualification is not possible, legacy waste can be qualified by spot checking according to the materials composition requirements The integration of the control of the material composition in the quality control system for radioactive waste is discussed and examples of control measures are given. With the materials-list and the packaging-list the Federal Office for Radiation Protection (BfS) provides an appropriate tool to describe the materials composition of radioactive waste packages. The control measures with respect to the materials composition integrate well in the established quality control framework for radioactive waste. The system is flexible enough to deal with waste products of different qualities: raw waste, qualified conditioned waste or legacy waste. Control measures to verify the materials composition can be accomplished with minimal radiation exposure and without undue burden on the waste producers and conditioners. (authors)

  17. Management of solid wastes in the iron and steel industry

    SciTech Connect

    El-Gohary, F.; El-khouly, M.S.

    1983-03-01

    Wastes from a local iron and steel factory operations are agglomeration of iron ore and sintering, pig iron manufacture, steel making, rolling mill operations, and pickling. Liquid slag, produced in the blast furnace, is granulated in water and used as a concrete additive. Other wastes are directed separately to sedimentation tanks. The settleable solids are reused, and the treated effluents are pumped to a cooling tower for recycling. As a result of the new manufacturing expansion, existing waste treatment facilities are not adequate, and it was found necessary to provide additional treatment techniques. Departmental, as well as composite wastes were treated using plain sedimentation, centrifugal sedimentation, or chemical coagulation, or a combination of these methods. The results obtained showed that the use of the hydrocyclone for solid-liquid separation is much more efficient than plain sedimentation. When this process was followed by coagulation, very promising results were obtained. The use of pickling liquor as a coagulant gave comparable results with alum and ferric chloride.

  18. Mixed waste chemical compatibility: A testing program for plastic packaging components

    SciTech Connect

    Nigrey, P.J.

    1995-12-01

    The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the United States have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). The design requirements for both hazardous [49 CFR 173.24 (e)(1)] and radioactive [49 CFR 173.412 (g)] materials packaging specify packaging compatibility, i.e., that the materials of the packaging @d any contents be chemically compatible with each other. Furthermore, Type A [49 CFR 173.412 (g)] and Type B (10 CFR 71.43) packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program attempts to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. This program has been described in considerable detail in an internal SNL document, the Chemical Compatibility Test Plan & Procedure Report (Nigrey 1993).

  19. Retail colour stability of lamb meat is influenced by breed type, muscle, packaging and iron concentration.

    PubMed

    Warner, R D; Kearney, G; Hopkins, D L; Jacob, R H

    2017-01-17

    The longissmus lumborum (LL) and semimembranosus (SM) muscles from 391 lamb carcasses, derived from various breed types, were used to investigate the effect of animal/muscle factors, packaging type [over-wrap (OW) or high oxygen modified atmosphere packaging (MAPO2)] and duration of display on redness of meat during simulated retail display. Using statistical models the time required (in days) for redness to reach a threshold value of 3.5 (below this is unacceptable) was predicted. High levels of iron in the SM, but not LL, reduced the time for redness to reach 3.5 by 2-2.6days in MAPO2 and 0.5-0.8days in OW. The greater the proportion of Merino breed type, the shorter was the time for redness to reach the value of 3.5, an effect consistent across muscles and packaging types. In summary, breed type, packaging format, muscle and muscle iron levels had a significant impact on colour stability of sheep meat in oxygen-available packaging systems.

  20. Hydrothermal carbonization of food waste and associated packaging materials for energy source generation.

    PubMed

    Li, Liang; Diederick, Ryan; Flora, Joseph R V; Berge, Nicole D

    2013-11-01

    Hydrothermal carbonization (HTC) is a thermal conversion technique that converts food wastes and associated packaging materials to a valuable, energy-rich resource. Food waste collected from local restaurants was carbonized over time at different temperatures (225, 250 and 275°C) and solids concentrations to determine how process conditions influence carbonization product properties and composition. Experiments were also conducted to determine the influence of packaging material on food waste carbonization. Results indicate the majority of initial carbon remains integrated within the solid-phase at the solids concentrations and reaction temperatures evaluated. Initial solids concentration influences carbon distribution because of increased compound solubilization, while changes in reaction temperature imparted little change on carbon distribution. The presence of packaging materials significantly influences the energy content of the recovered solids. As the proportion of packaging materials increase, the energy content of recovered solids decreases because of the low energetic retention associated with the packaging materials. HTC results in net positive energy balances at all conditions, except at a 5% (dry wt.) solids concentration. Carbonization of food waste and associated packaging materials also results in net positive balances, but energy needs for solids post-processing are significant. Advantages associated with carbonization are not fully realized when only evaluating process energetics. A more detailed life cycle assessment is needed for a more complete comparison of processes.

  1. A Fruit of Yucca Mountain: The Remote Waste Package Closure System

    SciTech Connect

    Kevin Skinner; Greg Housley; Colleen Shelton-Davis

    2011-11-01

    Was the death of the Yucca Mountain repository the fate of a technical lemon or a political lemon? Without caution, this debate could lure us away from capitalizing on the fruits of the project. In March 2009, Idaho National Laboratory (INL) successfully demonstrated the Waste Package Closure System, a full-scale prototype system for closing waste packages that were to be entombed in the now abandoned Yucca Mountain repository. This article describes the system, which INL designed and built, to weld the closure lids on the waste packages, nondestructively examine the welds using four different techniques, repair the welds if necessary, mitigate crack initiating stresses in the surfaces of the welds, evacuate and backfill the packages with an inert gas, and perform all of these tasks remotely. As a nation, we now have a proven method for securely sealing nuclear waste packages for long term storage—regardless of whether or not the future destination for these packages will be an underground repository. Additionally, many of the system’s features and concepts may benefit other remote nuclear applications.

  2. Packaging waste recycling in Europe: is the industry paying for it?

    PubMed

    da Cruz, Nuno Ferreira; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

    2014-02-01

    This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry. However, local governments are generally in charge of waste management, particularly in countries with Green Dot schemes or similar extended producer responsibility systems. This leads to the need of establishing a system of financial transfers between the industry and the local governments (particularly regarding the extra costs involved with selective collection and sorting). Using the same methodological approach, the authors also compare the costs and benefits of recycling from the perspective of local public authorities for France, Portugal and Romania. Since the purpose of the current paper is to take note of who is paying for the incremental costs of recycling and whether the industry (i.e. the consumer) is paying for the net financial costs of packaging waste management, environmental impacts are not included in the analysis. The work carried out in this paper highlights some aspects that are prone to be improved and raises several questions that will require further research. In the three countries analyzed more closely in this paper the industry is not paying the net financial cost of packaging waste management. In fact, if the savings attained by diverting packaging waste from other treatment (e.g. landfilling) and the public subsidies to the investment on the "recycling system" are not considered, it seems that the industry should increase the financial support to local authorities (by 125% in France, 50% in Portugal and 170% in Romania). However, in France and

  3. Packaging waste recycling in Europe: Is the industry paying for it?

    SciTech Connect

    Ferreira da Cruz, Nuno Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

    2014-02-15

    Highlights: • We study the recycling schemes of France, Germany, Portugal, Romania and the UK. • The costs and benefits of recycling are compared for France, Portugal and Romania. • The balance of costs and benefits depend on the perspective (strictly financial/economic). • Financial supports to local authorities ought to promote cost-efficiency. - Abstract: This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry. However, local governments are generally in charge of waste management, particularly in countries with Green Dot schemes or similar extended producer responsibility systems. This leads to the need of establishing a system of financial transfers between the industry and the local governments (particularly regarding the extra costs involved with selective collection and sorting). Using the same methodological approach, the authors also compare the costs and benefits of recycling from the perspective of local public authorities for France, Portugal and Romania. Since the purpose of the current paper is to take note of who is paying for the incremental costs of recycling and whether the industry (i.e. the consumer) is paying for the net financial costs of packaging waste management, environmental impacts are not included in the analysis. The work carried out in this paper highlights some aspects that are prone to be improved and raises several questions that will require further research. In the three countries analyzed more closely in this paper the industry is not paying the net financial cost of packaging waste

  4. Determination of activation energy of pyrolysis of carton packaging wastes and its pure components using thermogravimetry.

    PubMed

    Alvarenga, Larissa M; Xavier, Thiago P; Barrozo, Marcos Antonio S; Bacelos, Marcelo S; Lira, Taisa S

    2016-07-01

    Many processes have been used for recycling of carton packaging wastes. The pyrolysis highlights as a promising technology to be used for recovering the aluminum from polyethylene and generating products with high heating value. In this paper, a study on pyrolysis reactions of carton packaging wastes and its pure components was performed in order to estimate the kinetic parameters of these reactions. For this, dynamic thermogravimetric analyses were carried out and two different kinds of kinetic models were used: the isoconversional and Independent Parallel Reactions. Isoconversional models allowed to calculate the overall activation energy of the pyrolysis reaction, in according to their conversions. The IPR model, in turn, allowed the calculation of kinetic parameters of each one of the carton packaging and paperboard subcomponents. The carton packaging pyrolysis follows three separated stages of devolatilization. The first step is moisture loss. The second stage is perfectly correlated to devolatilization of cardboard. The third step is correlated to devolatilization of polyethylene.

  5. Vertical Drop of the Naval SNF Long Waste Package On Unyielding Surface

    SciTech Connect

    S. Mastilovic

    2006-07-21

    The purpose of this calculation is to determine the structural response of a Naval SNF (Spent Nuclear Fuel) Long Waste Package (WP) subjected to 2 m-vertical drop on unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design; calculation is performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element calculation is performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The result of this calculation is provided in terms of maximum stress intensities.

  6. 40 CFR 721.10667 - Slimes and sludges, aluminum and iron casting, wastewater treatment, solid waste.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... casting, wastewater treatment, solid waste. 721.10667 Section 721.10667 Protection of Environment... iron casting, wastewater treatment, solid waste. (a) Chemical substance and significant new uses... and iron casting, wastewater treatment, solid waste (PMN P-12-560; CAS No. 1391739-82-4;...

  7. 40 CFR 721.10667 - Slimes and sludges, aluminum and iron casting, wastewater treatment, solid waste.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... casting, wastewater treatment, solid waste. 721.10667 Section 721.10667 Protection of Environment... iron casting, wastewater treatment, solid waste. (a) Chemical substance and significant new uses... and iron casting, wastewater treatment, solid waste (PMN P-12-560; CAS No. 1391739-82-4;...

  8. Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

    SciTech Connect

    Adkins, H.E.

    1996-10-29

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete- Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

  9. Scale-up considerations relevant to experimental studies of nuclear waste-package behavior

    SciTech Connect

    Coles, D.G.; Peters, R.D.

    1986-04-01

    Results from a study that investigated whether testing large-scale nuclear waste-package assemblages was technically warranted are reported. It was recognized that the majority of the investigations for predicting waste-package performance to date have relied primarily on laboratory-scale experimentation. However, methods for the successful extrapolation of the results from such experiments, both geometrically and over time, to actual repository conditions have not been well defined. Because a well-developed scaling technology exists in the chemical-engineering discipline, it was presupposed that much of this technology could be applicable to the prediction of waste-package performance. A review of existing literature documented numerous examples where a consideration of scaling technology was important. It was concluded that much of the existing scale-up technology is applicable to the prediction of waste-package performance for both size and time extrapolations and that conducting scale-up studies may be technically merited. However, the applicability for investigating the complex chemical interactions needs further development. It was recognized that the complexity of the system, and the long time periods involved, renders a completely theoretical approach to performance prediction almost hopeless. However, a theoretical and experimental study was defined for investigating heat and fluid flow. It was concluded that conducting scale-up modeling and experimentation for waste-package performance predictions is possible using existing technology. A sequential series of scaling studies, both theoretical and experimental, will be required to formulate size and time extrapolations of waste-package performance.

  10. Iron Phosphate Glass-Containing Hanford Waste Simulant

    SciTech Connect

    Sevigny, Gary J.; Kimura, Marcia L.; Fischer, Christopher M.; Schweiger, M. J.; Rodriguez, Carmen P.; Kim, Dong-Sang; Riley, Brian J.

    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 is 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.

  11. Iron Phosphate Glass-Containing Hanford Waste Simulant

    SciTech Connect

    Sevigny, Gary J.; Kimura, Marcia L.; Fischer, Christopher M.; Schweiger, Michael J.; Kim, Dong-Sang

    2011-08-01

    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 is high in sulfates (4.37 wt% SO3). The primary objective of the test was to develop data to support a cost-benefit analysis as 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, and Mo-Sci Corporation.

  12. PROBABILISTIC ANALYSES OF WASTE PACKAGE QUANTITIES IMPACTED BY POTENTIAL IGNEOUS DISRUPTION AT YUCCA MOUNTAIN

    SciTech Connect

    M.G. Wallace

    2005-08-26

    A probabilistic analysis was conducted to estimate ranges for the numbers of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analyses include disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application for the Yucca Mountain Project (YMP). The first scenario, igneous intrusion, investigated the case where one or more igneous dikes intersect the repository. A swarm of dikes was characterized by distributions of length, width, azimuth, and number of dikes and the spacings between them. Through the use in part of a latin hypercube simulator and a modified video game engine, mathematical relationships were built between those parameters and the number of waste packages hit. Corresponding cumulative distribution function curves (CDFs) for the number of waste packages hit under several different scenarios were calculated. Variations in dike thickness ranges, as well as in repository magma bulkhead positions were examined through sensitivity studies. It was assumed that all waste packages in an emplacement drift would be impacted if that drift were intersected by a dike. Over 10,000 individual simulations were performed. Based on these calculations, out of a total of over 11,000 planned waste packages distributed over an area of approximately 5.5 km{sup 2} , the median number of waste packages impacted was roughly 1/10 of the total. Individual cases ranged from 0 waste packages to the entire inventory being impacted. The igneous intrusion analysis involved an explicit characterization of dike-drift intersections, built upon various distributions that reflect the uncertainties associated with the inputs. The second igneous scenario, volcanic eruption (eruptive conduits), considered the effects of conduits formed

  13. WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

    SciTech Connect

    J.W. Davis

    1996-07-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so.

  14. Probablistic Analyses of Waste Package Quantities Impacted by Potential Igneous Disruption at Yucca Mountain

    NASA Astrophysics Data System (ADS)

    Wallace, M. G.; Iuzzolina, H.

    2005-12-01

    A probabilistic analysis was conducted to estimate ranges for the numbers of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analysis includes disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application for the Yucca Mountain Project (YMP). The first scenario, igneous intrusion, investigated the case where one or more igneous dikes intersect the repository. A swarm of dikes was characterized by distributions of length, width, azimuth, and number of dikes and the spacings between them. Through the use in part of a latin hypercube simulator and a modified video game engine, mathematical relationships were built between those parameters and the number of waste packages hit. Corresponding cumulative distribution function curves (CDFs) for the number of waste packages hit under several different scenarios were calculated. Variations in dike thickness ranges, as well as in repository magma bulkhead positions were examined through sensitivity studies. It was assumed that all waste packages in an emplacement drift would be impacted if that drift was intersected by a dike. Over 10,000 individual simulations were performed. Based on these calculations, out of a total of over 11,000 planned waste packages distributed over an area of approximately 5.5 km2 , the median number of waste packages impacted was roughly 1/10 of the total. Individual cases ranged from 0 waste packages to the entire inventory being impacted. The igneous intrusion analysis involved an explicit characterization of dike-drift intersections, built upon various distributions that reflect the uncertainties associated with the inputs. The second igneous scenario, volcanic eruption (eruptive conduits), considered the effects of conduits formed in

  15. Concept for waste package environment tests in the Yucca Mountain exploratory shaft

    SciTech Connect

    Yow, J.L. Jr.

    1985-05-01

    The Nevada Nuclear Waste Storage Investigations (NNWSI) project is studying a tuffaceous rock unit located at Yucca Mountain on the western boundary of the Nevada Test Site, Nye County, Nevada. The objective is to evaluate the suitability of the volcanic rocks located above the water table at Yucca Mountain as a potential location for a repository for high level radioactive waste. As part of the NNWSI project, Lawrence Livermore National Laboratory is responsible for the design of the waste package and for determining the expected performance of the waste package in the repository environment. To design an optimal waste package system for the unsaturated emplacement environment, the mechanisms by which liquid water can return to contact the metal canister after peaking of the thermal load must be established. Definition of these flux and flow mechanisms is essential for estimating canister corrosion modes and rates. Therefore, three waste package environment tests are being designed for the in situ phase of exploratory shaft testing. These tests emphasize measurement techniques that offer the possibility of characterizing the movement of water into and through the pores and fractures of the densely welded Topopah Spring Member. Other measurement techniques will be used to examine the interactions between moisture migration and the thermomechanical rock mass behavior. Three reduced-scale heater tests will use electrical resistive heaters in a horizontal configuration. All three tests are designed to investigate moisture conditions in the rock during heating and cooling phases of a thermal cycle so that the effects of these moisture conditions on the performance of the waste package system may be established. 28 refs., 4 figs., 3 tabs.

  16. Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste

    SciTech Connect

    Wurm, K.J.; Miller, N.E.

    1982-11-01

    This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

  17. Natural additives and agricultural wastes in biopolymer formulations for food packaging

    NASA Astrophysics Data System (ADS)

    Valdés, Arantzazu; Mellinas, Ana Cristina; Ramos, Marina; Garrigós, María Carmen; Jiménez, Alfonso

    2014-02-01

    The main directions in food packaging research are targeted towards improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of new materials, and particularly innovative biopolymer formulations, can help to address these requirements and also with other packaging functions such as: food protection and preservation, marketing and smart communication to consumers. The use of biocomposites for active food packaging is one of the most studied approaches in the last years on materials in contact with food. Applications of these innovative biocomposites could help to provide new food packaging materials with improved mechanical, barrier, antioxidant and antimicrobial properties. From the food industry standpoint, concerns such as the safety and risk associated with these new additives, migration properties and possible human ingestion and regulations need to be considered. The latest innovations in the use of these innovative formulations to obtain biocomposites are reported in this review. Legislative issues related to the use of natural additives and agricultural wastes in food packaging systems are also discussed.

  18. Natural additives and agricultural wastes in biopolymer formulations for food packaging

    PubMed Central

    Valdés, Arantzazu; Mellinas, Ana Cristina; Ramos, Marina; Garrigós, María Carmen; Jiménez, Alfonso

    2014-01-01

    The main directions in food packaging research are targeted toward improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of new materials, and particularly innovative biopolymer formulations, can help to address these requirements and also with other packaging functions such as: food protection and preservation, marketing and smart communication to consumers. The use of biocomposites for active food packaging is one of the most studied approaches in the last years on materials in contact with food. Applications of these innovative biocomposites could help to provide new food packaging materials with improved mechanical, barrier, antioxidant, and antimicrobial properties. From the food industry standpoint, concerns such as the safety and risk associated with these new additives, migration properties and possible human ingestion and regulations need to be considered. The latest innovations in the use of these innovative formulations to obtain biocomposites are reported in this review. Legislative issues related to the use of natural additives and agricultural wastes in food packaging systems are also discussed. PMID:24790975

  19. Natural additives and agricultural wastes in biopolymer formulations for food packaging.

    PubMed

    Valdés, Arantzazu; Mellinas, Ana Cristina; Ramos, Marina; Garrigós, María Carmen; Jiménez, Alfonso

    2014-01-01

    The main directions in food packaging research are targeted toward improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of new materials, and particularly innovative biopolymer formulations, can help to address these requirements and also with other packaging functions such as: food protection and preservation, marketing and smart communication to consumers. The use of biocomposites for active food packaging is one of the most studied approaches in the last years on materials in contact with food. Applications of these innovative biocomposites could help to provide new food packaging materials with improved mechanical, barrier, antioxidant, and antimicrobial properties. From the food industry standpoint, concerns such as the safety and risk associated with these new additives, migration properties and possible human ingestion and regulations need to be considered. The latest innovations in the use of these innovative formulations to obtain biocomposites are reported in this review. Legislative issues related to the use of natural additives and agricultural wastes in food packaging systems are also discussed.

  20. Potential vertical movement of large heat-generating waste packages in salt.

    SciTech Connect

    Clayton, Daniel James; Martinez, Mario J.; Hardin, Ernest.

    2013-05-01

    With renewed interest in disposal of heat-generating waste in bedded or domal salt formations, scoping analyses were conducted to estimate rates of waste package vertical movement. Vertical movement is found to result from thermal expansion, from upward creep or heave of the near-field salt, and from downward buoyant forces on the waste package. A two-pronged analysis approach was used, with thermal-mechanical creep modeling, and coupled thermal-viscous flow modeling. The thermal-mechanical approach used well-studied salt constitutive models, while the thermal-viscous approach represented the salt as a highly viscous fluid. The Sierra suite of coupled simulation codes was used for both approaches. The waste package in all simulations was a right-circular cylinder with the density of steel, in horizontal orientation. A time-decaying heat generation function was used to represent commercial spent fuel with typical burnup and 50-year age. Results from the thermal-mechanical base case showed approximately 27 cm initial uplift of the package, followed by gradual relaxation closely following the calculated temperature history. A similar displacement history was obtained with the package density set equal to that of salt. The slight difference in these runs is attributable to buoyant displacement (sinking) and is on the order of 1 mm in 2,000 years. Without heat generation the displacement stabilizes at a fraction of millimeter after a few hundred years. Results from thermal-viscous model were similar, except that the rate of sinking was constant after cooldown, at approximately 0.15 mm per 1,000 yr. In summary, all calculations showed vertical movement on the order of 1 mm or less in 2,000 yr, including calculations using well-established constitutive models for temperature-dependent salt deformation. Based on this finding, displacement of waste packages in a salt repository is not a significant repository performance issue.

  1. Evaluation and compilation of DOE waste package test data: Biannual report, August 1986-January 1987

    SciTech Connect

    Interrante, C.; Escalante, E.; Fraker, A.; Harrison, S.; Shull, R.; Linzer, M.; Ricker, R.; Ruspi, J.

    1987-10-01

    This report summarizes results of the National Bureau of Standards (NBS) evaluations of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon and stainless steels, and copper. In the section on tuff, the current level of understanding of several canister materials is questioned. Within the Basalt Waste Isolation Project (BWIP) section, discussions are given on problems concerning groundwater, materials for use in the metallic overpack, and diffusion through the packing. For the proposed salt site, questions are raised on the work on both ASTM A216 Steel and Ti-Code 12. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) is covered. NBS reviews of selected DOE technical reports and a summary of current waste-package activities of the Materials Characterization Center (MCC) is presented. Using a database management system, a computerized database for storage and retrieval of reviews and evaluations of HLW data has been developed and is described. 17 refs., 2 figs., 2 tabs.

  2. Chemical compatibility screening results of plastic packaging to mixed waste simulants

    SciTech Connect

    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 specific 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.

  3. Safety evaluation for packaging (onsite) depleted uranium waste boxes

    SciTech Connect

    McCormick, W.A.

    1997-08-27

    This safety evaluation for packaging (SEP) allows the one-time shipment of ten metal boxes and one wooden box containing depleted uranium material from the Fast Flux Test Facility to the burial grounds in the 200 West Area for disposal. This SEP provides the analyses and operational controls necessary to demonstrate that the shipment will be safe for the onsite worker and the public.

  4. Summary strategy for compliance with postclosure requirements for the waste package for the Salt Repository Project: Final report

    SciTech Connect

    Not Available

    1988-03-01

    This document presents a summary of the strategy of the Salt Repository Project (SRP) to show compliance with the requirements for the waste package after permanent closure of the repository at the site in Deaf Smith County, Texas. The postclosure requirements that govern the performance of the waste package are those in 10 CFR 60.113 for substantially complete containment of the waste and for gradual release of radionuclides after the containment period, and for the postclosure design requirements in 10 CFR 60.135. Also, the waste package plays a role in showing compliance with the total system release requirement in 40 CFR 191.13. 12 refs.

  5. Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment

    SciTech Connect

    Pierce, Eric M.; McGrail, B. Peter; Rodriguez, Elsa A.; Schaef, Herbert T.; Saripalli, Prasad; Serne, R. Jeffrey; Krupka, Kenneth M.; Martin, P. F.; Baum, Steven R.; Geiszler, Keith N.; Reed, Lunde R.; Shaw, Wendy J.

    2004-09-01

    This data package documents the experimentally derived input data on the representative waste glasses; LAWA44, LAWB45, and LAWC22. This data will be used for Subsurface Transport Over Reactive Multi-phases (STORM) simulations of the Integrated Disposal Facility (IDF) for immobilized low-activity waste (ILAW). The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in July 2005. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali (Na+)-hydrogen (H+) ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow (PUF) and product consistency (PCT) tests where used for accelerated weathering or aging of the glasses in order to determine a chemical reaction network of secondary phases that form. The majority of the thermodynamic data used in this data package were extracted from the thermody-namic database package shipped with the geochemical code EQ3/6, version 8.0. Because of the expected importance of 129I release from secondary waste streams being sent to IDF from various thermal treatment processes, parameter estimates for diffusional release and solubility-controlled release from cementitious waste forms were estimated from the available literature.

  6. Design of a nuclear-waste package for emplacement in tuff

    SciTech Connect

    O`Neal, W.C.; Rothman, A.J.; Gregg, D.W.; Hockman, J.N.; Revelli, M.A.; Russell, E.W.; Schornhorst, J.R.

    1983-02-01

    Design, modeling, and testing activities are under way at LLNL in the development of high level nuclear waste package designs. We discuss the geological characteristics affecting design, the 10CFR60 design requirements, conceptual designs, metals for containment barriers, economic analysis, thermal modeling, and performance modeling.

  7. Effect of aged waste package and basalt on radioelement release

    SciTech Connect

    Seitz, M.G.; Vandegrift, G.F.; Bowers, D.L.; Gerding, T.J.

    1983-01-01

    Results of experiments are described that combine backfill, radioactive waste, and repository host rock in a single flowing groundwater stream in a manner analogous to a hydraulic breach of a waste repository. The experimental design is used to identify the chemical interactions that would occur if repository components were breached by flowing water. The results indicate that of three parameters studied, the alteration of the repository components as might occur upon aging had the most substantial influence on the migration of radioactive elements dissolved from the solid radioactive waste. The other two parameters, the metal alloy used in the apparatus and an ionizing radiation field imposed on the experimental apparatus, had little or no measurable effect on radioactive element transport by flowing water. Inasmuch as the alteration of the repository materials represent aging in an actual repository, it is concluded that changes with age may detrimentally affect the ability of a repository to isolate plutonium and neptunium, and possibly other radioactive elements in nuclear waste. 37 references, 2 figures, 2 tables.

  8. Using Single-Camera 3-D Imaging to Guide Material Handling Robots in a Nuclear Waste Package Closure System

    SciTech Connect

    Rodney M. Shurtliff

    2005-09-01

    Nuclear reactors for generating energy and conducting research have been in operation for more than 50 years, and spent nuclear fuel and associated high-level waste have accumulated in temporary storage. Preparing this spent fuel and nuclear waste for safe and permanent storage in a geological repository involves developing a robotic packaging system—a system that can accommodate waste packages of various sizes and high levels of nuclear radiation. During repository operation, commercial and government-owned spent nuclear fuel and high-level waste will be loaded into casks and shipped to the repository, where these materials will be transferred from the casks into a waste package, sealed, and placed into an underground facility. The waste packages range from 12 to 20 feet in height and four and a half to seven feet in diameter. Closure operations include sealing the waste package and all its associated functions, such as welding lids onto the container, filling the inner container with an inert gas, performing nondestructive examinations on welds, and conducting stress mitigation. The Idaho National Laboratory is designing and constructing a prototype Waste Package Closure System (WPCS). Control of the automated material handling is an important part of the overall design. Waste package lids, welding equipment, and other tools must be moved in and around the closure cell during the closure process. These objects are typically moved from tool racks to a specific position on the waste package to perform a specific function. Periodically, these objects are moved from a tool rack or the waste package to the adjacent glovebox for repair or maintenance. Locating and attaching to these objects with the remote handling system, a gantry robot, in a loosely fixtured environment is necessary for the operation of the closure cell. Reliably directing the remote handling system to pick and place the closure cell equipment within the cell is the major challenge.

  9. Effects of Hanford tank simulant waste on plastic packaging to components

    SciTech Connect

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

    1995-12-01

    We have developed a chemical compatibility program for the evaluation of plastic packaging components which may be incorporated in packaging for transporting mixed waste forms. Consistent with the methodology outlined in this paper, we have performed the second phase of this experimental program to determine the effects of simulant Hanford Tank mixed wastes on packaging materials. This effort involved the comprehensive testing of five plastic liner materials in the aqueous mixed waste simulant. The testing protocol involved exposing the respective materials to {approximately}1, 3, 6, and 40 kGy of gamma radiation followed by 7, 14, 28, 180 day exposures to the waste simulant at 18, 50, and 60{degree}C. From the limited data analyses performed to date in this study, we have identified the fluorocarbon Kel-F{trademark} as having the greatest chemical compatibility after having been exposed to 40 kGy gamma radiation followed by exposure to the Hanford Tank simulant mixed waste at 60{degree}C. The most stricking observation from this study was the poor performance of Teflon under these conditions.

  10. Petrologic and geochemical characterization of the Bullfrog Member of the Crater Flat Tuff: outcrop samples used in waste package experiments

    SciTech Connect

    Knauss, K.G.

    1983-09-01

    In support of the Waste Package Task within the Nevada Nuclear Waste Storage Investigation (NNWSI), experiments on hydrothermal rock/water interaction, corrosion, thermomechanics, and geochemical modeling calculations are being conducted. All of these activities require characterization of the initial bulk composition, mineralogy, and individual phase geochemistry of the potential repository host rock. This report summarizes the characterization done on samples of the Bullfrog Member of the Crater Flat Tuff (Tcfb) used for Waste Package experimental programs. 11 references, 17 figures, 3 tables.

  11. Potential Biogenic Corrosion of Alloy 22, A Candidate Nuclear Waste Packaging Materials, Under Simulated Repository Conditions

    SciTech Connect

    Horn, J.M.; Martin, S.I.; Rivera, A.J.; Bedrossian, P.J.; Lian, T.

    2000-01-12

    The U.S. Department of Energy has been charged with assessing the suitability of a geologic nuclear waste repository at Yucca Mountain (YM), NV. Microorganisms, both those endogenous to the repository site and those introduced as a result of construction and operational activities, may contribute to the corrosion of metal nuclear waste packaging and thereby decrease their useful lifetime as barrier materials. Evaluation of potential Microbiological Influenced Corrosion (MIC) on candidate waste package materials was undertaken reactor systems incorporating the primary elements of the repository: YM rock (either non-sterile or presterilized), material coupons, and a continual feed of simulated YM groundwater. Periodically, both aqueous reactor efflux and material coupons were analyzed for chemical and surfacial characterization. Alloy 22 coupons exposed for a year at room temperature in reactors containing non-sterile YM rock demonstrated accretion of chromium oxide and silaceous scales, with what appear to be underlying areas of corrosion.

  12. STRUCTURAL CALCULATION OF AN EMPLACEMENT PALLET STATICALLY LOADED BY A WASTE PACKAGE

    SciTech Connect

    S. Mastilovic

    2000-02-02

    The purpose of this calculation is to determine the structural response of the emplacement pallet (EP) subjected to static load from the mounted waste package (WP). The scope of this document is limited to reporting the calculation results in terms of stress intensity magnitudes. This calculation is associated with the waste emplacement systems design; calculations are performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element solutions are performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The results of these calculations are provided in terms of maximum stress intensity magnitudes.

  13. Demands placed on waste package performance testing and modeling by some general results on reliability analysis

    SciTech Connect

    Chesnut, D.A.

    1991-09-01

    Waste packages for a US nuclear waste repository are required to provide reasonable assurance of maintaining substantially complete containment of radionuclides for 300 to 1000 years after closure. The waiting time to failure for complex failure processes affecting engineered or manufactured systems is often found to be an exponentially-distributed random variable. Assuming that this simple distribution can be used to describe the behavior of a hypothetical single barrier waste package, calculations presented in this paper show that the mean time to failure (the only parameter needed to completely specify an exponential distribution) would have to be more than 10{sub 7} years in order to provide reasonable assurance of meeting this requirement. With two independent barriers, each would need to have a mean time to failure of only 10{sup 5} years to provide the same reliability. Other examples illustrate how multiple barriers can provide a strategy for not only achieving but demonstrating regulatory compliance.

  14. Chemical Environment at Waste Package Surfaces in a High-Level Radioactive Waste Repository

    SciTech Connect

    Carroll, S; Alai, M; Craig, L; Gdowski, G; Hailey, P; Nguyen, Q A; Rard, J; Staggs, K; Sutton, M; Wolery, T

    2005-05-26

    We have conducted a series of deliquescence, boiling point, chemical transformation, and evaporation experiments to determine the composition of waters likely to contact waste package surfaces over the thermal history of the repository as it heats up and cools back down to ambient conditions. In the above-boiling period, brines will be characterized by high nitrate to chloride ratios that are stable to higher temperatures than previously predicted. This is clearly shown for the NaCl-KNO{sub 3} salt system in the deliquescence and boiling point experiments in this report. Our results show that additional thermodynamic data are needed in nitrate systems to accurately predict brine stability and composition due to salt deliquescence in dust deposited on waste package surfaces. Current YMP models capture dry-out conditions but not composition for NaCl-KNO{sub 3} brines, and they fail to predict dry-out conditions for NaCl-KNO{sub 3}-NaNO{sub 3} brines. Boiling point and deliquescence experiments are needed in NaCl-KNO{sub 3}-NaNO{sub 3} and NaCl-KNO{sub 3}-NaNO{sub 3}-Ca(NO{sub 3}){sub 2} systems to directly determine dry-out conditions and composition, because these salt mixtures are also predicted to control brine composition in the above-boiling period. Corrosion experiments are needed in high temperature and high NO{sub 3}:Cl brines to determine if nitrate inhibits corrosion in these concentrated brines at temperatures above 160 C. Chemical transformations appear to be important for pure calcium- and magnesium-chloride brines at temperatures greater than 120 C. This stems from a lack of acid gas volatility in NaCl/KNO{sub 3} based brines and by slow CO{sub 2}(g) diffusion in alkaline brines. This suggests that YMP corrosion models based on bulk solution experiments over the appropriate composition, temperature, and relative humidity range can be used to predict corrosion in thin brine films formed by salt deliquescence. In contrast to the above-boiling period, the

  15. Disposal of LLW and ILW in Germany - Characterisation and Documentation of Waste Packages with Respect to the Change of Requirements

    SciTech Connect

    Bandt, G.; Spicher, G.; Steyer, St.; Brennecke, P.

    2008-07-01

    Since the 1998 termination of LLW and ILW emplacement in the Morsleben repository (ERAM), Germany, the treatment, conditioning and documentation of radioactive waste products and packages have been continued on the basis of the waste acceptance requirements as of 1995, prepared for the Konrad repository near Salzgitter in Lower Saxony, Germany. The resulting waste products and packages are stored in interim storage facilities. Due to the Konrad license issued in 2002 the waste acceptance requirements have to be completed by additional requirements imposed by the licensing authority, e. g. for the declaration of chemical waste package constituents. Therefore, documentation of waste products and packages which are checked by independent experts and are in parts approved by the responsible authority (Office for Radiation Protection, BfS) up to now will have to be checked again for fulfilling the final waste acceptance requirements prior to disposal. In order to simplify these additional checks, databases are used to ensure an easy access to all known facts about the waste packages. A short balance of the existing waste products and packages which are already checked and partly approved by BfS as well as an overview on the established databases ensuring a fast access to the known facts about the conditioning processes is presented. (authors)

  16. Effect of ionizing radiation on the waste package environment

    SciTech Connect

    Reed, D.T.; Van Konynenburg, R.A.

    1991-05-01

    The radiolytic production of nitrogen oxides, nitrogen acids and ammonia are discussed in relation to the expected environment in a high-level waste repository that may be constructed at the Yucca Mountain site if it is found to be suitable. Both literature data and repository-relevant data are summarized for air-water vapor systems. The limiting cases of a dry air and a pure water vapor gas phase are also discussed. Design guidelines and recommendations, based solely on the potential consequence of radiation enhancement of corrosion, are given. 13 refs., 5 figs., 1 tab.

  17. Pyrolysis of plastic packaging waste: A comparison of plastic residuals from material recovery facilities with simulated plastic waste

    SciTech Connect

    Adrados, A.

    2012-05-15

    Highlights: Black-Right-Pointing-Pointer Pyrolysis of plastic waste. Black-Right-Pointing-Pointer Comparison of different samples: real waste, simulated and real waste + catalyst. Black-Right-Pointing-Pointer Study of the effects of inorganic components in the pyrolysis products. - Abstract: Pyrolysis may be an alternative for the reclamation of rejected streams of waste from sorting plants where packing and packaging plastic waste is separated and classified. These rejected streams consist of many different materials (e.g., polyethylene (PE), polypropylene (PP), polystyrene (PS), polyvinyl chloride (PVC), polyethylene terephthalate (PET), acrylonitrile butadiene styrene (ABS), aluminum, tetra-brik, and film) for which an attempt at complete separation is not technically possible or economically viable, and they are typically sent to landfills or incinerators. For this study, a simulated plastic mixture and a real waste sample from a sorting plant were pyrolyzed using a non-stirred semi-batch reactor. Red mud, a byproduct of the aluminum industry, was used as a catalyst. Despite the fact that the samples had a similar volume of material, there were noteworthy differences in the pyrolysis yields. The real waste sample resulted, after pyrolysis, in higher gas and solid yields and consequently produced less liquid. There were also significant differences noted in the compositions of the compared pyrolysis products.

  18. Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218

    SciTech Connect

    French, David M.; Hayes, Timothy A.; Pope, Howard L.; Enriquez, Alejandro E.; Carson, Peter H.

    2013-07-01

    In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely productive Waste

  19. Induction motors with minimum waste of magnetic iron

    NASA Astrophysics Data System (ADS)

    Boyko, Y. P.; Makarov, F. K.; Kremeshnyy, Y. A.; Stepanyants, E. A.

    1984-11-01

    An induction motor with minimum waste of magnetic iron has been developed. The stator core is made of silicon-iron steel strip, its width equal to the core height, and wound on edge to form a stack. Stator slots closed on the bore side and open on the outside for winding are formed together with teeth separating them by corrugating a silicon-iron steel strip of width equal to the stack length and sliding it, after winding, inside the core with an insulation interlayer between them. The thickness of this interlayer does not exceed 20% of the air gap between stator and rotor, but even this degrades motor starting and running performance. The performance is also degraded by leakage of magnetic flux through the slot bridge on the bore side and by nonuniformity of the air gap width resulting from corrugation of the stator bore surface. These drawbacks are partly compensated by lining the slots with film insulation instead of sleeves or tubes, thus making more slot space available for copper. Prototypes were built with 0.5 mm thick strip of 2212 steel for the teeth and strip of 08KP steel for the core, separated by a 0.03 mm layer of EKD-14 epoxy compound, with the air gap correspondingly narrowed from 0.25 to 0.2 mm. The stator winding is a singlelayer one of enamelled and triple vinyl-covered 0.67/0.73 mm copper wire, the slots insulated with Dacron film.

  20. Prompt gamma neutron activation analysis of toxic elements in radioactive waste packages.

    PubMed

    Ma, J-L; Carasco, C; Perot, B; Mauerhofer, E; Kettler, J; Havenith, A

    2012-07-01

    The French Alternative Energies and Atomic Energy Commission (CEA) and National Radioactive Waste Management Agency (ANDRA) are conducting an R&D program to improve the characterization of long-lived and medium activity (LL-MA) radioactive waste packages. In particular, the amount of toxic elements present in radioactive waste packages must be assessed before they can be accepted in repository facilities in order to avoid pollution of underground water reserves. To this aim, the Nuclear Measurement Laboratory of CEA-Cadarache has started to study the performances of Prompt Gamma Neutron Activation Analysis (PGNAA) for elements showing large capture cross sections such as mercury, cadmium, boron, and chromium. This paper reports a comparison between Monte Carlo calculations performed with the MCNPX computer code using the ENDF/B-VII.0 library and experimental gamma rays measured in the REGAIN PGNAA cell with small samples of nickel, lead, cadmium, arsenic, antimony, chromium, magnesium, zinc, boron, and lithium to verify the validity of a numerical model and gamma-ray production data. The measurement of a ∼20kg test sample of concrete containing toxic elements has also been performed, in collaboration with Forschungszentrum Jülich, to validate the model in view of future performance studies for dense and large LL-MA waste packages.

  1. Response of elastomeric packaging components to a corrosive simulant mixed waste

    SciTech Connect

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

    1997-10-01

    The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the US have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation and the US Nuclear Regulatory Commission. Based on these national requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at SNL. In this paper, the authors present the results of Part B of the second phase of this testing program. The first phase screened five liner materials and six seal materials towards four simulant mixed wastes. Part A of the second phase involved the comprehensive testing of five candidate liner materials to an aqueous Hanford Tank simulant mixed waste. Part B involved similar testing on elastomeric materials, ethylene-propylene and butadiene-acrylonitrile rubber. The comprehensive testing protocol involved exposing the respective materials to a matrix of four gamma radiation doses ({approximately}1, 3, 6, and 40 kGy), three temperatures (18, 50, and 60 C), and four exposure times (7, 14, 28, and 180 days). Following their exposure to these combinations of conditions, the materials were evaluated by measuring six material properties. These properties were specific gravity, dimensional changes, hardness, vapor transport rates, compression set, and mechanical properties.

  2. A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

    SciTech Connect

    Brown, G. S.; Cashwell, J. W.; Apple, M. L.

    1991-01-01

    Shipments of radioactive material (RAM) constitute but a small fraction of the total hazardous materials shipped in the United States each year. Public perception, however, of the potential consequences of a release from a transportation package containing RAM has resulted in significant regulation of transport operations, both to ensure the integrity of a package in accident conditions and to place operational constraints on the shipper. Much of this attention has focused on shipments of spent nuclear fuel and high level wastes which, although comprising a very small number of total shipments, constitute a majority of the total curies transported on an annual basis. This report discusses the shipment of these highly radioactive materials.

  3. Packaging and environment in Europe (EUREKA PACK-EE): Automated sorting of aluminum from domestic waste

    SciTech Connect

    Bertaud, Y.; Guillermet, R.; Lemaire, H.; Cael, J.; Nijhof, G.; Rossel, H.

    1996-10-01

    A large European research project named EUREKA PACK-EE (for Packaging Environment in Europe) has been started in June 1993. The project carries out research and innovation into collection, sorting and valorization of types of packaging, with improvements to sorting procedures capable of meeting the needs of the recovery and recycling industries. The project includes studies of consumer behavior concerning waste, which will influence the choice of sorting techniques. The results are presented for automated sorting by different eddy current (EC) engineering and in two other papers for thermal separation and purification.

  4. A Robust Power Remote Manipulator for Use in Waste Sorting, Processing, and Packaging - 12158

    SciTech Connect

    Cole, Matt; Martin, Scott

    2012-07-01

    Disposition of radioactive waste is one of the Department of Energy's (DOE's) highest priorities. A critical component of the waste disposition strategy is shipment of Transuranic (TRU) waste from DOE's Oak Ridge Reservation to the Waste Isolation Plant Project (WIPP) in Carlsbad, New Mexico. This is the mission of the DOE TRU Waste Processing Center (TWPC). The remote-handled TRU waste at the Oak Ridge Reservation is currently in a mixed waste form that must be repackaged in to meet WIPP Waste Acceptance Criteria (WAC). Because this remote-handled legacy waste is very diverse, sorting, size reducing, and packaging will require equipment flexibility and strength that is not possible with standard master-slave manipulators. To perform the wide range of tasks necessary with such diverse, highly contaminated material, TWPC worked with S.A. Technology (SAT) to modify SAT's Power Remote Manipulator (PRM) technology to provide the processing center with an added degree of dexterity and high load handling capability inside its shielded cells. TWPC and SAT incorporated innovative technologies into the PRM design to better suit the operations required at TWPC, and to increase the overall capability of the PRM system. Improving on an already proven PRM system will ensure that TWPC gains the capabilities necessary to efficiently complete its TRU waste disposition mission. The collaborative effort between TWPC and S.A. Technology has yielded an extremely capable and robust solution to perform the wide range of tasks necessary to repackage TRU waste containers at TWPC. Incorporating innovative technologies into a proven manipulator system, these PRMs are expected to be an important addition to the capabilities available to shielded cell operators. The PRMs provide operators with the ability to reach anywhere in the cell, lift heavy objects, perform size reduction associated with the disposition of noncompliant waste. Factory acceptance testing of the TWPC Powered Remote

  5. Application of fluidization to separate packaging waste plastics.

    PubMed

    Carvalho, M Teresa; Ferreira, Célia; Portela, Antía; Santos, João Tiago

    2009-03-01

    The objective of the experimental work described in this paper is the study of the separation of PS (polystyrene) from PET (polyethylene terephthalate) and PVC (polyvinyl chloride) from drop-off points using a fluidized bed separator. This is a low-cost process commonly used in the hydro-classification of mineral ores. Firstly, experimental tests were carried out with artificial granulated samples with different grain sizes, types and sources of plastic ("separability tests"). The particle settling velocities were determined under different operating conditions. Then, based on the results, the laboratory tests continued with real mixtures of waste plastics ("separation tests") and the efficiency of the process was evaluated. From a PET-rich mixture, a concentrate of PS with a 75% grade in PS was produced while the underflow was quite clear from PS (grade less than 0.5% in PS).

  6. Thermal testing of packages for transport of radioactive wastes

    SciTech Connect

    Koski, J.A.

    1994-12-31

    Shipping containers for radioactive materials must be shown capable of surviving tests specified by regulations such as Title 10, Code of Federal Regulations, Part 71 (called 10CFR71 in this paper) within the United States. Equivalent regulations hold for other countries such as Safety Series 6 issued by the International Atomic Energy Agency. The containers must be shown to be capable of surviving, in order, drop tests, puncture tests, and thermal tests. Immersion testing in water is also required, but must be demonstrated for undamaged packages. The thermal test is intended to simulate a 30 minute exposure to a fully engulfing pool fire that could occur if a transport accident involved the spill of large quantities of hydrocarbon fuels. Various qualification methods ranging from pure analysis to actual pool fire tests have been used to prove regulatory compliance. The purpose of this paper is to consider the alternatives for thermal testing, point out the strengths and weaknesses of each approach, and to provide the designer with the information necessary to make informed decisions on the proper test program for the particular shipping container under consideration. While thermal analysis is an alternative to physical testing, actual testing is often emphasized by regulators, and this report concentrates on these testing alternatives.

  7. Gravity packaging final waste recovery based on gravity separation and chemical imaging control.

    PubMed

    Bonifazi, Giuseppe; Serranti, Silvia; Potenza, Fabio; Luciani, Valentina; Di Maio, Francesco

    2017-02-01

    Plastic polymers are characterized by a high calorific value. Post-consumer plastic waste can be thus considered, in many cases, as a typical secondary solid fuels according to the European Commission directive on End of Waste (EoW). In Europe the practice of incineration is considered one of the solutions for waste disposal waste, for energy recovery and, as a consequence, for the reduction of waste sent to landfill. A full characterization of these products represents the first step to profitably and correctly utilize them. Several techniques have been investigated in this paper in order to separate and characterize post-consumer plastic packaging waste fulfilling the previous goals, that is: gravity separation (i.e. Reflux Classifier), FT-IR spectroscopy, NIR HyperSpectralImaging (HSI) based techniques and calorimetric test. The study demonstrated as the proposed separation technique and the HyperSpectral NIR Imaging approach allow to separate and recognize the different polymers (i.e. PolyVinyl Chloride (PVC), PolyStyrene (PS), PolyEthylene (PE), PoliEtilene Tereftalato (PET), PolyPropylene (PP)) in order to maximize the removal of the PVC fraction from plastic waste and to perform the full quality control of the resulting products, can be profitably utilized to set up analytical/control strategies finalized to obtain a low content of PVC in the final Solid Recovered Fuel (SRF), thus enhancing SRF quality, increasing its value and reducing the "final waste".

  8. Characterization of dolochar wastes generated by the sponge iron industry

    NASA Astrophysics Data System (ADS)

    Dwari, Ranjan Kumar; Rao, Danda Srinivas; Swar, Akhila Kumar; Reddy, Palli Sita Ram; Mishra, Barada Kanta

    2012-11-01

    Solid wastes generated by the metallurgical industry contribute significantly towards the enhancement of environmental pollution. The handling, utilization, and safe disposal of these solid wastes are major concerns for the world. Dolochar is such a solid waste generated by the sponge iron industry. Investigations were carried out on the physical, mineralogical, and chemical characteristics for the efficient utilization of dolochar. The detailed studies on physico-chemical properties and petrography were carried out by optical microscopy, X-ray diffraction (XRD), and scanning electron microscopy (SEM). Characterization studies revealed that the dolochar consists of quartz (free as well as locked), free lime, Fe particles, and Ca or Mg and/or Ca+Mg+Fe oxide phases. The washability data of -300 μm dolochar samples indicated that clean coal with 41wt% ash at 18% yield can be produced from dolochar with 78wt% ash. The studies further suggested that the liberation of the dolochar is hard to achieve for clear separation. The dolochar is observed to have high ash fusion temperature and the unburned carbon can be best utilized for power generation.

  9. Modeling of Stress Corrosion Cracking for High Level Radioactive-Waste Packages

    SciTech Connect

    Lu, S C; Gordon, G M; Andresen, P L; Herrera, M L

    2003-06-20

    A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain radioactive-waste repository. SCC is one form of environmentally assisted cracking due to three factors, which must be present simultaneously: metallurgical susceptibility, critical environment, and static (or sustained) tensile stresses. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is Alloy 22, a highly corrosion resistant alloy, the environment is represented by the water film present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the stress is principally the weld induced residual stress. SCC has historically been separated into ''initiation'' and ''propagation'' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding). To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulas for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, the time to through-wall penetration for the waste package can be calculated. The SDFR model relates the advance (or propagation) of cracks, subsequent to the crack initiation from bare metal surface, to the metal oxidation transients that occur when the protective film at the crack tip is continually ruptured and repassivated. A crack, however, may reach the ''arrest'' state before it enters the ''propagation'' phase. There exists a threshold stress intensity factor, which provides a criterion for determining if an initiated crack or pre-existing manufacturing flaw will reach the ''arrest'' state. This paper presents the research

  10. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    SciTech Connect

    J.P. Nicot

    2000-09-29

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the WP. This

  11. DIFFERENCES AND SIMILARITIES IN ANDRA'S ASSESSMENT OF ACTIVITIES CARRIED OUT BY RADIOACTIVE WASTE GENERATORS AND AFFECTING THE QUALITY OF IL-LL SHORT-LIVED WASTE PACKAGES AND HL-IL LONG-LIVED WASTE PACKAGES

    SciTech Connect

    Trentesaux, C.; Cairon, P.; Dumont, J.-N.; Felix, B.; Losada, F.

    2003-02-27

    In both cases of packages for either low-level and intermediate-level short-lived (LL-IL/SL) or high-level and intermediate-level long-lived (HL-IL/LL) radioactive waste, Andra has defined a quality reference system, manages it, follows up its appropriate implementation in production plants and verifies its effectiveness in production. The purpose of such a reference system is to ensure, in the first case, that waste packages comply with the Centre de l'Aube's acceptance criteria and, in the second case, that the characteristics submitted by the waste generators to Andra as input data for the deep geological repository project reflect the actual production conditions. In that context, the three management steps of the quality reference system include differences due to the fact that HL-IL/SL packages have not been submitted yet to any technical acceptance criterion. Compliance with any such criterion should be the subject of a characterization report during the qualification phase and of a examination during the verification phase. The management of the quality reference system also involves similarities that facilitate the joint work carried out by Andra with the waste generators, especially in the facilities where both package types are produced.

  12. Evaluation and compilation of DOE waste package test data: Biannual report, August 1987--January 1988

    SciTech Connect

    Interrante, C.; Escalante, E.; Fraker, A.; Ondik, H.; Plante, E.; Ricker, R.; Ruspi, J.

    1988-08-01

    This report summarizes results of the National Bureau of Standards (NBS) evaluations on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Since enactment of the Budget Reconciliation Act for Fiscal Year 1988, the Yucca Mountain, Nevada, site (in which tuff is the geologic medium) is the only site that will be characterized for use as high-level nuclear waste repository. During the reporting period of August 1987 to January 1988, five reviews were completed for tuff, and these were grouped into the categories: ferrous alloys, copper, groundwater chemistry, and glass. Two issues are identified for the Yucca Mountain site: the approach used to calculate corrosion rates for ferrous alloys, and crevice corrosion was observed in a copper-nickel alloy. Plutonium can form pseudo-colloids that may facilitate transport. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) and activities of the DOE Materials Characterization Center (MCC) for the 6-month reporting period are also included. 27 refs., 3 figs.

  13. Review of DOE Waste Package Program. Semiannual report, October 1984-March 1985. Volume 8

    SciTech Connect

    Davis, M.S.

    1985-12-01

    A large number of technical reports on waste package component performance were reviewed over the last year in support of the NRC`s review of the Department of Energy`s (DOE`s) Environmental Assessment reports. The intent was to assess in some detail the quantity and quality of the DOE data and their relevance to the high-level waste repository site selection process. A representative selection of the reviews is presented for the salt, basalt, and tuff repository projects. Areas for future research have been outlined. 141 refs.

  14. DOE Waste Package Project. Quarterly progress report, October 1, 1993--December 31, 1993

    SciTech Connect

    Ladkany, S.G.

    1993-12-31

    The Waste Package Project research activities continued in all research areas. The areas include: Container structural and stress analysis; Nuclear fission criticality studies; Investigation of canister design concepts and corrosion studies; Heat transfer studies; Fluid flow in porous media and radionuclide transport in near field rock; Studies of stresses and stability of the rock formations resulting from the thermal loading of the fuel elements and the multi tunnel concept being analyzed; Characterization of a Faulted Rock Tunnel Model Using Photoelastic and Finite Element Studies; Experiment studies of the dynamic response of a flexible three-link robot using strain gages and Lagrange polynomials; and Robotic manipulation of the nuclear waste container.

  15. Hanford low-level waste process chemistry testing data package

    SciTech Connect

    Smith, H.D.; Tracey, E.M.; Darab, J.G.; Smith, P.A.

    1996-03-01

    Recently, the Tri-Party Agreement (TPA) among the State of Washington Department of Ecology, U.S. Department of Energy (DOE) and the US Environmental Protection Agency (EPA) for the cleanup of the Hanford Site was renegotiated. The revised agreement specifies vitrification as the encapsulation technology for low level waste (LLW). A demonstration, testing, and evaluation program underway at Westinghouse Hanford Company to identify the best overall melter-system technology available for vitrification of Hanford Site LLW to meet the TPA milestones. Phase I is a {open_quotes}proof of principle{close_quotes} test to demonstrate that a melter system can process a simulated highly alkaline, high nitrate/nitrite content aqueous LLW feed into a glass product of consistent quality. Seven melter vendors were selected for the Phase I evaluation: joule-heated melters from GTS Duratek, Incorporated (GDI); Envitco, Incorporated (EVI); Penberthy Electomelt, Incorporated (PEI); and Vectra Technologies, Incorporated (VTI); a gas-fired cyclone burner from Babcock & Wilcox (BCW); a plasma torch-fired, cupola furnace from Westinghouse Science and Technology Center (WSTC); and an electric arc furnace with top-entering vertical carbon electrodes from the U.S. Bureau of Mines (USBM).

  16. FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation

    SciTech Connect

    K. Mon

    2004-10-11

    The purpose of this report is to evaluate and document the inclusion or exclusion of features, events and processes (FEPs) with respect to drip shield and waste package modeling used to support the Total System Performance Assessment for License Application (TSPA-LA). Thirty-three FEPs associated with the waste package and drip shield performance have been identified (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). A screening decision, either ''included'' or ''excluded,'' has been assigned to each FEP, with the technical bases for screening decisions, as required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, and f) [DIRS 156605]. The FEPs analyses in this report address issues related to the degradation and potential failure of the drip shield and waste package over the post closure regulatory period of 10,000 years after permanent closure. For included FEPs, this report summarizes the disposition of the FEP in TSPA-LA. For excluded FEPs, this report provides the technical bases for the screening arguments for exclusion from TSPA-LA. The analyses are for the TSPA-LA base-case design (BSC 2004 [DIRS 168489]), where a drip shield is placed over the waste package without backfill over the drip shield (BSC 2004 [DIRS 168489]). Each FEP includes one or more specific issues, collectively described by a FEP name and description. The FEP description encompasses a single feature, event, or process, or a few closely related or coupled processes, provided the entire FEP can be addressed by a single specific screening argument or TSPA-LA disposition. The FEPs were assigned to associated Project reports, so the screening decisions reside with the relevant subject-matter experts.

  17. Petrologic and geochemical characterization of the Topopah Spring Member of the Paintbrush Tuff: outcrop samples used in waste package experiments

    SciTech Connect

    Knauss, K.G.

    1984-06-01

    This report summarizes characterization studies conducted with outcrop samples of Topopah Spring Member of the Paintbrush Tuff (Tpt). In support of the Waste Package Task within the Nevada Nuclear Waste Storage Investigation (NNWSI), Tpt is being studied both as a primary object and as a constituent used to condition water that will be reacted with waste form, canister, or packing material. These studies directly or indirectly support NNWSI subtasks concerned with waste package design and geochemical modeling. To interpret the results of subtask experiments, it is necessary to know the exact nature of the starting material in terms of the intial bulk composition, mineralogy, and individual phase geochemistry. 31 figures, 5 tables.

  18. An econometric analysis of regional differences in household waste collection: The case of plastic packaging waste in Sweden

    SciTech Connect

    Hage, Olle Soederholm, Patrik

    2008-07-01

    The Swedish producer responsibility ordinance mandates producers to collect and recycle packaging materials. This paper investigates the main determinants of collection rates of household plastic packaging waste in Swedish municipalities. This is done by the use of a regression analysis based on cross-sectional data for 252 Swedish municipalities. The results suggest that local policies, geographic/demographic variables, socio-economic factors and environmental preferences all help explain inter-municipality collection rates. For instance, the collection rate appears to be positively affected by increases in the unemployment rate, the share of private houses, and the presence of immigrants (unless newly arrived) in the municipality. The impacts of distance to recycling industry, urbanization rate and population density on collection outcomes turn out, though, to be both statistically and economically insignificant. A reasonable explanation for this is that the monetary compensation from the material companies to the collection entrepreneurs vary depending on region and is typically higher in high-cost regions. This implies that the plastic packaging collection in Sweden may be cost ineffective. Finally, the analysis also shows that municipalities that employ weight-based waste management fees generally experience higher collection rates than those municipalities in which flat and/or volume-based fees are used.

  19. An econometric analysis of regional differences in household waste collection: the case of plastic packaging waste in Sweden.

    PubMed

    Hage, Olle; Söderholm, Patrik

    2008-01-01

    The Swedish producer responsibility ordinance mandates producers to collect and recycle packaging materials. This paper investigates the main determinants of collection rates of household plastic packaging waste in Swedish municipalities. This is done by the use of a regression analysis based on cross-sectional data for 252 Swedish municipalities. The results suggest that local policies, geographic/demographic variables, socio-economic factors and environmental preferences all help explain inter-municipality collection rates. For instance, the collection rate appears to be positively affected by increases in the unemployment rate, the share of private houses, and the presence of immigrants (unless newly arrived) in the municipality. The impacts of distance to recycling industry, urbanization rate and population density on collection outcomes turn out, though, to be both statistically and economically insignificant. A reasonable explanation for this is that the monetary compensation from the material companies to the collection entrepreneurs vary depending on region and is typically higher in high-cost regions. This implies that the plastic packaging collection in Sweden may be cost ineffective. Finally, the analysis also shows that municipalities that employ weight-based waste management fees generally experience higher collection rates than those municipalities in which flat and/or volume-based fees are used.

  20. Review of waste package verification tests. Semiannual report, April 1985-September 1985

    SciTech Connect

    Soo, P.

    1986-01-01

    Several studies were completed this period to evaluate experimental and analytical methodologies being used in the DOE waste package program. The first involves a determination of the relevance of the test conditions being used by DOE to characterize waste package component behavior in a salt repository system. Another study focuses on the testing conditions and procedures used to measure radionuclide solubility and colloid formation in repository groundwaters. An attempt was also made to evaluate the adequacy of selected waste package performance codes. However, the latter work was limited by an inability to obtain several codes from DOE. Nevertheless, it was possible to comment briefly on the structures and intents of the codes based on publications in the open literature. The final study involved an experimental program to determine the likelihood of stress-corrosion cracking of austenitic stainless steels and Incoloy 825 in simulated tuff repository environments. Tests for six-month exposure periods in water and air-steam conditions are described. 52 figs., 48 tabs.

  1. Magma flow and interaction with waste packages in a geologic repository at Yucca Mountain, Nevada

    NASA Astrophysics Data System (ADS)

    Marsh, Bruce D.; Coleman, Neil M.

    2009-05-01

    The likelihood that a nuclear waste repository at Yucca Mountain could be intersected by igneous activity is very low, but its potential consequences are nevertheless important to performance assessments. An ongoing critical area of concern is the nature and magnitude of the thermal interaction of magma with tunnel walls, high-level nuclear waste packages, and waste forms. Previous assessments consider a variety of dynamic scenarios, but large uncertainties remain in understanding the rheological nature of the magma likely to be involved and its behavior within a repository drift. Here we specifically address the issue of magma rheology during degassing, cooling, and solidification as basaltic magma approaches Earth's surface and enters a drift. Magma containing significant amounts (> ~ 2 wt.%) of dissolved water (Wet Magma), as is anticipated for this region, at or near its liquidus temperature and saturated with water at 200 MPa is at a temperature near or below the 1-atm solidus temperature. Isentropic ascent from this near liquidus temperature promotes extensive solidification and/or glassification. Exsolving water with approach to the surface promotes rapid vesiculation leading to fragmentation and tephra production. With continued ascent the still water-saturated magma traverses the solidification phase field and undergoes a combination of rapid crystallization and quenching, becoming a glassy highly viscous (~ 10 8 Pa s) mass of greatly reduced mobility. This immobility is reflected in the high effective viscosity regulating flows from nearby cinder cones associated with wet basalt. This also matches well with the experimentally determined rheology of dry basalt glass. This rheology greatly restricts the mobility of basalt within repository drifts, amounting to < 10 m per day. Magma in this state quenches rapidly (~ 10 cm/min) on waste packages. Wet basalt is explosive, but relatively immobile as lava. Dry Magma is not explosive, but highly mobile as lava

  2. Development of backfill material as an engineered barrier in the waste package system- Interim topical report

    SciTech Connect

    Wheelwright, E.J.; Hodges, F.N.; Bray, L.A.; Westsik, J.H. Jr.; Lester, D.H.; Nakai, T.L.; Spaeth, M.E.; Stula, R.T.

    1981-09-01

    A backfill barrier, emplaced between the containerized waste and the host rock, can both protect the other engineered barriers and act as a primary barrier to the release of radionuclides from the waste package. Attributes that a backfill should provide in order to carry out its required function have been identified. Primary attributes are those that have a direct effect upon the release and transport of radionuclides from the waste package. Supportive attributes do not directly affect radionuclide release but are necessary to support the primary attributes. The primary attributes, in order of importance, are: minimize (retard or exclude) the migration of ground water between the host rock and the waste canister system; retard the migration of selected chemical species (corrosive species and radionuclides) in the ground water; control the Eh and pH of the ground water within the waste-package environment. The supportive attributes are: self-seal any cracks or discontinuities in the backfill or interfacing host geology; retain performance properties at all repository temperatures; retain peformance properties during and after receiving repository levels of gamma radiation; conduct heat from the canister system to the host geology; retain mechanical properties and provide resistance to applied mechanical forces; retain morphological stability and compatibility with structural barriers and with the host geology for required period of time. Screening and selection of candidate backfill materials has resulted in a preliminary list of materials for testing. Primary emphasis has been placed on sodium and calcium bentonites and zeolites used in conjunction with quartz sand or crushed host rock. Preliminary laboratory studies have concentrated on permeability, sorption, swelling pressure, and compaction properties of candidate backfill materials.

  3. The Role of Capillary Barrier in Reducing Moisture Content on Waste Packages

    SciTech Connect

    N.D. Rosenberg; T.A. Buscheck; D. Wildenschild; Y. Sun

    1999-10-29

    Assessment of the performance of engineered capillary barriers at the potential Yucca Mountain nuclear waste repository site, in which 1.67-m-diameter waste packages are to be emplaced in 5-m-diameter tunnels according to current design, brings up aspects not commonly considered in more typical applications of capillary barriers (e.g., near-surface landfills). Engineered capillary barriers typically consist of two layers of granular materials with a sloping interface, in which the contrast in capillarity between the layers keeps infiltrating water in the upper layer. One issue is the effect of thermohydrologic processes that would occur at elevated repository temperatures (and temperature gradients). For example, backfill materials may be altered from that of the as-placed material by the hydrothermal regime imposed by the emplacement of waste in the repository, changing hydrologic properties in a way that degrades the performance of the barrier. A reduction of permeability in the upper layer might diminish the capacity of the upper layer to divert incoming seepage or to cause a ''vapor lid'' whereby buoyant vapor flow would be trapped, then condense and drain onto waste packages. Other concerns are the result of highly spatially and temporally variable seepage distribution and the very limited spatial scale available for flow attenuation and diversion.

  4. The unlikelihood of localized corrosion of nuclear waste packages arising from deliquescent brine formation

    NASA Astrophysics Data System (ADS)

    Apted, M.; Arthur, R.; King, F.; Langmuir, D.; Kessler, J.

    2005-01-01

    The Nuclear Waste Technical Review Board (NWTRB) recently postulated a scenario for the formation of a deliquescent, divalent-cation (Ca, Mg) chloride brine from wind-blown dust on the surface of an alloy 22 container designed to hold radioactive waste. The NWTRB suggested that such brines could lead to localized penetrations of waste packages at a proposed repository in Yucca Mountain, Nevada. In response, the Electric Power Research Institute (EPRI) sponsored an independent analysis, specifically examining and responding to a series of decision points in the NWTRB’s scenario. Only if all of these questions could be answered affirmatively would NWTRB’s scenario result in potential noncompliance with regulatory criteria. The EPRI analysis found that none of the questions has an affirmative answer.

  5. Aspiration requirements for the transportation of retrievably stored waste in the TRUPACT-2 package

    SciTech Connect

    Djordjevic, S.; Drez, P.; Murthy, D. ); Temus, C. )

    1990-01-01

    The Transuranic Package Transporter-II (TRUPACT-II) is the shipping package to be used for the transportation of contact-handled transuranic (CH TRU) waste between the various US Department of Energy (DOE) sites, and to the Waste Isolation Pilot Plant (WIPP) located near Carlsbad, New Mexico. Waste (payload) containers to be transported in the TRUPACT-II package are required to be vented prior to being shipped. Venting'' refers to the installation of one or more carbon composite filters in the lid of the container, and the puncturing of a rigid liner (if present). This ensures that there is no buildup of pressure or potentially flammable gas concentrations in the container prior to transport. Payload containers in retrievable storage that have been stored in an unvented condition at the DOE sites, may have generated and accumulated potentially flammable concentrations of gases (primarily due to generation of hydrogen by radiolysis) during the unvented storage period. Such payload containers need to be aspirated for a sufficient period of time until safe pre-transport conditions (acceptably low hydrogen concentrations) are achieved. The period of time for which a payload container needs to be in a vented condition before qualifying for transport in a TRUPACT-II package is defined as the aspiration time.'' This paper presents the basis for evaluating the minimum aspiration time for a payload container that has been in unvented storage. Three different options available to the DOE sites for meeting the aspiration requirements are described in this paper. 4 refs., 2 figs.

  6. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    DOEpatents

    Boatner, Lynn A.; Sales, Brian C.

    1989-01-01

    Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

  7. Pyrolysis of plastic packaging waste: A comparison of plastic residuals from material recovery facilities with simulated plastic waste.

    PubMed

    Adrados, A; de Marco, I; Caballero, B M; López, A; Laresgoiti, M F; Torres, A

    2012-05-01

    Pyrolysis may be an alternative for the reclamation of rejected streams of waste from sorting plants where packing and packaging plastic waste is separated and classified. These rejected streams consist of many different materials (e.g., polyethylene (PE), polypropylene (PP), polystyrene (PS), polyvinyl chloride (PVC), polyethylene terephthalate (PET), acrylonitrile butadiene styrene (ABS), aluminum, tetra-brik, and film) for which an attempt at complete separation is not technically possible or economically viable, and they are typically sent to landfills or incinerators. For this study, a simulated plastic mixture and a real waste sample from a sorting plant were pyrolyzed using a non-stirred semi-batch reactor. Red mud, a byproduct of the aluminum industry, was used as a catalyst. Despite the fact that the samples had a similar volume of material, there were noteworthy differences in the pyrolysis yields. The real waste sample resulted, after pyrolysis, in higher gas and solid yields and consequently produced less liquid. There were also significant differences noted in the compositions of the compared pyrolysis products.

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

    SciTech Connect

    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 liner 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.

  9. Reuse of waste iron as a partial replacement of sand in concrete.

    PubMed

    Ismail, Zainab Z; Al-Hashmi, Enas A

    2008-11-01

    One of the major environmental issues in Iraq is the large quantity of waste iron resulting from the industrial sector which is deposited in domestic waste and in landfills. A series of 109 experiments and 586 tests were carried out in this study to examine the feasibility of reusing this waste iron in concrete. Overall, 130 kg of waste iron were reused to partially replace sand at 10%, 15%, and 20% in a total of 1703 kg concrete mixtures. The tests performed to evaluate waste-iron concrete quality included slump, fresh density, dry density, compressive strength, and flexural strength tests: 115 cubes of concrete were molded for the compressive strength and dry density tests, and 87 prisms were cast for the flexural strength tests. This work applied 3, 7, 14, and 28 days curing ages for the concrete mixes. The results confirm that reuse of solid waste material offers an approach to solving the pollution problems that arise from an accumulation of waste in a production site; in the meantime modified properties are added to the concrete. The results show that the concrete mixes made with waste iron had higher compressive strengths and flexural strengths than the plain concrete mixes.

  10. Feasibility study of fissile mass quantification by photofission delayed gamma rays in radioactive waste packages using MCNPX

    NASA Astrophysics Data System (ADS)

    Simon, Eric; Jallu, Fanny; Pérot, Bertrand; Plumeri, Stéphane

    2016-12-01

    The feasibility of fissile mass quantification in large, long-lived medium activity radioactive waste packages using photofission delayed gamma rays has been assessed with MCNPX. The detection limit achievable is lower than the expected uranium mass in these waste packages, but the important sensibility to the waste matrix density and sample localization imposes to get an accurate measurement of these parameters. An isotope discrimination method based on gamma-ray ratios has been evaluated showing that photofission delayed gamma rays can be used to measure the fissile mass as well as the total uranium mass.

  11. Review of waste package verification tests. Semiannual report, October 1984-March 1985

    SciTech Connect

    Soo, P.

    1985-07-01

    The potential of WAPPA, a second-generation waste package system code, to meet the needs of the regulatory community is analyzed. The analysis includes an indepth review of WAPPA`s individual process models and a review of WAPPA`s operation. It is concluded that the code is of limited use to the NRC in the present form. Recommendations for future improvement, usage, and implementation of the code are given. This report also describes the results of a testing program undertaken to determine the chemical environment that will be present near a high-level waste package emplaced in a basalt repository. For this purpose, low carbon 1020 steel (a current BWIP reference container material), synthetic basaltic groundwater and a mixture of bentonite and basalt were exposed, in an autoclave, to expected conditions some period after repository sealing (150{sup 0}C, {approx_equal}10.4 MPa). Parameters measured include changes in gas pressure with time and gas composition, variation in dissolved oxygen (DO), pH and certain ionic concentrations of water in the packing material across an imposed thermal gradient, mineralogic alteration of the basalt/bentonite mixture, and carbon steel corrosion behavior. A second testing program was also initiated to check the likelihood of stress corrosion cracking of austenitic stainless steels and Incoloy 825 which are being considered for use as waste container materials in the tuff repository program. 82 refs., 70 figs., 27 tabs.

  12. Contaminant Release Data Package for Residual Waste in Single-Shell Hanford Tanks

    SciTech Connect

    Deutsch, William J.; Cantrell, Kirk J.; Krupka, Kenneth M.

    2007-12-01

    The Hanford Federal Facility Agreement and Consent Order requires that a Resource Conservation and Recovery Act (RCRA) Facility Investigation report be submitted to the Washington State Department of Ecology. The RCRA Facility Investigation report will provide a detailed description of the state of knowledge needed for tank farm performance assessments. This data package provides detailed technical information about contaminant release from closed single-shell tanks necessary to support the RCRA Facility Investigation report. It was prepared by Pacific Northwest National Laboratory (PNNL) for CH2M HILL Hanford Group, Inc., which is tasked by the U.S. Department of Energy (DOE) with tank closure. This data package is a compilation of contaminant release rate data for residual waste in the four Hanford single-shell tanks (SSTs) that have been tested (C-103, C-106, C-202, and C-203). The report describes the geochemical properties of the primary contaminants of interest from the perspective of long-term risk to groundwater (uranium, technetium-99, iodine-129, chromium, transuranics, and nitrate), the occurrence of these contaminants in the residual waste, release mechanisms from the solid waste to water infiltrating the tanks in the future, and the laboratory tests conducted to measure release rates.

  13. TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

    SciTech Connect

    NA

    2005-10-25

    The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair.

  14. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    SciTech Connect

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  15. Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and Waste Treatment, Storage and Disposal Activities

    SciTech Connect

    Jardine, L J; Borisov, G B

    2004-07-21

    A fifth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held February 16-18, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 46 Russian attendees from 14 different Russian organizations and six non-Russian attendees, four from the US and two from France. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C.

  16. Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material

    SciTech Connect

    G. Gordon

    2004-10-13

    Stress corrosion cracking is one of the most common corrosion-related causes for premature breach of metal structural components. Stress corrosion cracking is the initiation and propagation of cracks in structural components due to three factors that must be present simultaneously: metallurgical susceptibility, critical environment, and static (or sustained) tensile stresses. This report was prepared according to ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of this report is to provide an evaluation of the potential for stress corrosion cracking of the engineered barrier system components (i.e., the drip shield, waste package outer barrier, and waste package stainless steel inner structural cylinder) under exposure conditions consistent with the repository during the regulatory period of 10,000 years after permanent closure. For the drip shield and waste package outer barrier, the critical environment is conservatively taken as any aqueous environment contacting the metal surfaces. Appendix B of this report describes the development of the SCC-relevant seismic crack density model (SCDM). The consequence of a stress corrosion cracking breach of the drip shield, the waste package outer barrier, or the stainless steel inner structural cylinder material is the initiation and propagation of tight, sometimes branching, cracks that might be induced by the combination of an aggressive environment and various tensile stresses that can develop in the drip shields or the waste packages. The Stainless Steel Type 316 inner structural cylinder of the waste package is excluded from the stress corrosion cracking evaluation because the Total System Performance Assessment for License Application (TSPA-LA) does not take credit for the inner cylinder. This document provides a detailed description of the process-level models that can be applied to assess the performance of Alloy 22

  17. Production of pig iron from red mud waste fines using thermal plasma technology

    NASA Astrophysics Data System (ADS)

    Jayasankar, K.; Ray, P. K.; Chaubey, A. K.; Padhi, A.; Satapathy, B. K.; Mukherjee, P. S.

    2012-08-01

    Red mud, an insoluble residue produced during alkali leaching of bauxite, is considered as a low-grade iron ore containing 30% to 50% iron. The present paper deals with the use of thermal plasma technology for producing pig iron from red mud waste fines. The smelting reduction of red mud was carried out in a 35 kW DC extended arc thermal plasma reactor. Red mud was properly mixed with fluxes and graphite (fixed carbon, 99%) as a reductant as per stoichiometric requirement. The effect of various process parameters like a reductant, fluxes and smelting time on iron recovery was studied and optimized. An optimum condition for the maximum recovery of iron was obtained. A new thermal plasma process applicable to direct iron making from red mud waste fines that would achieve significant utilization of red mud was proposed.

  18. Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. Erratum

    SciTech Connect

    Smith, Gary L.

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  19. Dose Rate Calculations for the 2-MCO/2-DHLW Waste Package

    SciTech Connect

    G. Radulescu

    2000-10-03

    The objective of this calculation is to determine the dose rates on the external surfaces of the waste package (WP) containing two Hanford defense high-level waste (DHLW) glass canisters and two Hanford multi-canister overpacks (MCO). Each MCO is loaded with the N Reactor spent nuclear fuel (SNF). The information provided by the sketches attached to this calculation is that of the potential design for the WP type considered in this calculation. The scope of this calculation is limited to reporting dose rates averaged over segments of the WP radial and axial surfaces and of surfaces 1 m and 2 m from the WP. The results of this calculation will be used to assess the shielding performance of the 2-MC012-DHLW WP engineering design.

  20. Evaluation of performance indicators applied to a material recovery facility fed by mixed packaging waste.

    PubMed

    Mastellone, Maria Laura; Cremiato, Raffaele; Zaccariello, Lucio; Lotito, Roberta

    2017-03-14

    Most of the integrated systems for municipal solid waste management aim to increase the recycling of secondary materials by means of physical processes including sorting, shredding and reprocessing. Several restrictions prevent from reaching a very high material recycling efficiency: the variability of the composition of new-marketed materials used for packaging production and its shape and complexity are critical issues. The packaging goods are in fact made of different materials (aluminium, polymers, paper, etc.), possibly assembled, having different shape (flat, cylindrical, one-dimensional, etc.), density, colours, optical properties and so on. These aspects limit the effectiveness and efficiency of the sorting and reprocessing plants. The scope of this study was to evaluate the performance of a large scale Material Recovery Facility (MRF) by utilizing data collected during a long period of monitoring. The database resulted from the measured data has been organized in four sections: (1) data related to the amount and type of inlet waste; (2) amount and composition of output products and waste; (3) operating data (such as worked hours for shift, planned and unscheduled maintenance time, setting parameters of the equipment, and energy consumption for shift); (4) economic data (value of each product, disposal price for the produced waste, penalty for non-compliance of products and waste, etc.). A part of this database has been utilized to build an executive dashboard composed by a set of performance indicators suitable to measure the effectiveness and the efficiency of the MRF operations. The dashboard revealed itself as a powerful tool to support managers and engineers in their decisions in respect to the market demand or compliance regulation variation as well as in the designing of the lay-out improvements. The results indicated that the 40% of the input waste was recovered as valuable products and that a large part of these (88%) complied with the standards of

  1. A Review of Iron Phosphate Glasses and Recommendations for Vitrifying Hanford Waste

    SciTech Connect

    Delbert E. Ray; Chandra S. Ray

    2013-11-01

    This report contains a comprehensive review of the research conducted, world-wide, on iron phosphate glass over the past ~30 years. Special attention is devoted to those iron phosphate glass compositions which have been formulated for the purpose of vitrifying numerous types of nuclear waste, with special emphasis on the wastes stored in the underground tanks at Hanford WA. Data for the structural, chemical, and physical properties of iron phosphate waste forms are reviewed for the purpose of understanding their (a) outstanding chemical durability which meets all current DOE requirements, (b) high waste loadings which can exceed 40 wt% (up to 75 wt%) for several Hanford wastes, (c) low melting temperatures, can be as low as 900°C for certain wastes, and (d) high tolerance for “problem” waste components such as sulfates, halides, and heavy metals (chromium, actinides, noble metals, etc.). Several recommendations are given for actions that are necessary to smoothly integrate iron phosphate glass technology into the present waste treatment plans and vitrification facilities at Hanford.

  2. EVALUATION OF THE CORROSIVITY OF DUST DEPOSITED ON WASTE PACKAGES AT YUCCA MOUNTAIN, NEVADA

    SciTech Connect

    C. Bryan; R. Jack; T, Wolery; D. Shields; M. Sutton; E. Hardin; D. Barr

    2005-08-03

    Potentially corrosive brines can form during post-closure by deliquescence of salt minerals in dust deposited on the surface of waste packages at Yucca Mountain during operations and the pre-closure ventilation period. Although thermodynamic modeling and experimental studies of brine deliquescence indicates that brines are likely to form, they will be nitrate-rich and non-corrosive. Processes that modify the brines following deliquescence are beneficial with respect to inhibition of corrosion. For example, acid degassing (HCl, HNO{sub 3}) could dry out brines, but kinetic limitations are likely to limit the effect to increasing their passivity by raising the pH and increasing the NO{sub 3}/Cl ratio. Predicted dust quantities and maximum brine volumes on the waste package surface are small, and physical isolation of salt minerals in the dust may inhibit formation of eutectic brines and decrease brine volumes. If brines do contact the WP surface, small droplet volumes and layer thicknesses do not support development of diffusive gradients necessary for formation on separate anodic-cathodic zones required for localized corrosion. Finally, should localized corrosion initiate, corrosion product buildup will stifle corrosion, by limiting oxygen access to the metal surface, by capillary retention of brine in corrosion product porosity, or by consumption of brine components (Cl{sup -}).

  3. Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier

    SciTech Connect

    T. Wolery

    2005-02-22

    This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks.

  4. Evaluation and compilation of DOE waste package test data; Biannual report, August 1988--January 1989: Volume 6

    SciTech Connect

    Interrante, C.G.; Escalante, E.; Fraker, A.C.

    1990-11-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period August 1988 through January 1989. Included are reviews of related materials research and plans, activities for the DOE Materials Characterization Center, information on the Yucca Mountain Project, and other information regarding supporting research and special assistance. NIST comments are given on the Yucca Mountain Consultation Draft Site Characterization Plan (CDSCP) and on the Waste Compliance Plan for the West Valley Demonstration Project (WVDP) High-Level Waste (HLW) Form. 3 figs.

  5. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    SciTech Connect

    Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

    1985-10-01

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document.

  6. GLASS FEASIBILITY STUDY: VITRIFICATION OF OAK RIDGE NATIONAL LABORATORY GUNITE WASTE USING IRON PHOSPHATE GLASS (U)

    SciTech Connect

    Fellinger, T.

    1996-03-01

    This report describes the results of a glass feasibility study on vitrification of Oak Ridge National Laboratory (ORNL) Gunite waste into an Iron Phosphate glass. This glass feasibility study is part of a larger ORNL Gunite and Associated Tanks Treatability program (TTPSR1-6-WT-31). The treatability program explores different immobilization techniques of placing Gunite waste into a glass or grout form for long term storage. ORNL Gunite tanks contain waste that originated from years of various ORNL Research and Development programs. The available analyses of the Gunite Waste Tanks indicate, uranium and/or thorium as the dominant chemical constituent (50% +) and Cs{sup 137} the primary radionuclide. This information was utilized in determining a preliminary iron phosphate glass formulation. Chemical and physical properties: processing temperature, waste loading capability, chemical durability, density and redox were determined.

  7. Evaluating laser-driven Bremsstrahlung radiation sources for imaging and analysis of nuclear waste packages.

    PubMed

    Jones, Christopher P; Brenner, Ceri M; Stitt, Camilla A; Armstrong, Chris; Rusby, Dean R; Mirfayzi, Seyed R; Wilson, Lucy A; Alejo, Aarón; Ahmed, Hamad; Allott, Ric; Butler, Nicholas M H; Clarke, Robert J; Haddock, David; Hernandez-Gomez, Cristina; Higginson, Adam; Murphy, Christopher; Notley, Margaret; Paraskevoulakos, Charilaos; Jowsey, John; McKenna, Paul; Neely, David; Kar, Satya; Scott, Thomas B

    2016-11-15

    A small scale sample nuclear waste package, consisting of a 28mm diameter uranium penny encased in grout, was imaged by absorption contrast radiography using a single pulse exposure from an X-ray source driven by a high-power laser. The Vulcan laser was used to deliver a focused pulse of photons to a tantalum foil, in order to generate a bright burst of highly penetrating X-rays (with energy >500keV), with a source size of <0.5mm. BAS-TR and BAS-SR image plates were used for image capture, alongside a newly developed Thalium doped Caesium Iodide scintillator-based detector coupled to CCD chips. The uranium penny was clearly resolved to sub-mm accuracy over a 30cm(2) scan area from a single shot acquisition. In addition, neutron generation was demonstrated in situ with the X-ray beam, with a single shot, thus demonstrating the potential for multi-modal criticality testing of waste materials. This feasibility study successfully demonstrated non-destructive radiography of encapsulated, high density, nuclear material. With recent developments of high-power laser systems, to 10Hz operation, a laser-driven multi-modal beamline for waste monitoring applications is envisioned.

  8. 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.

  9. Physical and chemical characteristics of lead-iron phosphate nuclear waste glasses

    SciTech Connect

    Sales, B.C.; Boatner, L.A.

    1985-05-01

    Experimental determinations of the properties of lead-iron phosphate glasses pertinent to their application to the problem of permanently disposing of high-level nuclear wastes have been carried out. These investigations included studies of the composition and physical properties of nuclear waste glasses (NWG), as well as the effect of preparation conditions. Lead-iron phosphate nuclear waste glasses were prepared by dissolving simulated US defense wastes or simulated commercial power reactor wastes in molten lead-iron phosphate melts at temperatures between 900 and 1050/sup 0/C. The measured physical and chemical properties of the nuclear waste glasses formed by cooling these melts and annealing included the following: (1) aqueous corrosion resistance as a function of the solution pH, solution temperature, and glass composition, (2) glass density, (3) thermal expansion coefficient, (4) glass transition temperature and softening point, (5) heat capacity, (6) critical cooling rate, (7) temperature for the maximum crystallization rate, (8) relative solubility of waste oxides in the glass melt, (9) reactions between the molten glass and the melting crucible (Pt, ZrO/sub 2/, Al/sub 2/O/sub 3/), and (10 studies of possible metal cannister materials. Experimental results for the lead-iron phosphate NWG are compared to available data for borosilicate NWG. Relative to borosilicate NWG, the lead-iron phosphate glasses have several distinct advantages which include a much lower aqueous corrosion rate, a lower preparation temperature, and the ability to immobilize many types of commercial and defense-related high-level radioactive wastes. 34 refs., 18 figs., 10 tabs.

  10. An investigation of the usability of sound recognition for source separation of packaging wastes in reverse vending machines.

    PubMed

    Korucu, M Kemal; Kaplan, Özgür; Büyük, Osman; Güllü, M Kemal

    2016-10-01

    In this study, we investigate the usability of sound recognition for source separation of packaging wastes in reverse vending machines (RVMs). For this purpose, an experimental setup equipped with a sound recording mechanism was prepared. Packaging waste sounds generated by three physical impacts such as free falling, pneumatic hitting and hydraulic crushing were separately recorded using two different microphones. To classify the waste types and sizes based on sound features of the wastes, a support vector machine (SVM) and a hidden Markov model (HMM) based sound classification systems were developed. In the basic experimental setup in which only free falling impact type was considered, SVM and HMM systems provided 100% classification accuracy for both microphones. In the expanded experimental setup which includes all three impact types, material type classification accuracies were 96.5% for dynamic microphone and 97.7% for condenser microphone. When both the material type and the size of the wastes were classified, the accuracy was 88.6% for the microphones. The modeling studies indicated that hydraulic crushing impact type recordings were very noisy for an effective sound recognition application. In the detailed analysis of the recognition errors, it was observed that most of the errors occurred in the hitting impact type. According to the experimental results, it can be said that the proposed novel approach for the separation of packaging wastes could provide a high classification performance for RVMs.

  11. Preliminary Systems Design Study assessment report. Volume 6, Waste Isolation Pilot Plant and transportation package acceptable concepts

    SciTech Connect

    Mayberry, J.L.; Feizollahi, F.; Del Signore, J.C.

    1992-01-01

    The System Design Study (SDS), part of the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examined techniques for the remediation of hazardous and transuranic waste stored at Radioactive Waste Management Complex`s Subsurface Disposal Area at the INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. The SDS resulted in the development of technology requirements including demonstration, testing, and evaluation activities needed for implementing each concept. This volume contains introduction section containing a brief SDS background and lists the general assumptions and considerations used during the development of the system concepts. The introduction section is followed by sections describing two system concepts that produce a waste form in compliance with the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC) and transportation package (TRAMPAC) requirements. This system concept category is referred to as Waste Form 4, ``WIPP and TRAMPAC Acceptable.`` The following two system concepts are under this category: Sort, Treat, and Repackage System (4-BE-2); Volume Reduction and Packaging System (4-BE-4).

  12. Quarterly progress report on the DOE Waste Package project at the University of Nevada, Las Vegas, July 1, 1993 through September 30, 1993

    SciTech Connect

    Ladkany, S.G.

    1993-11-01

    Progress reports are presented for the following tasks: overview and progress of waste package project and container design; waste container design considerations (criticality analysis, experimental drift model); waste container alternate design considerations; thermal simulation of high level nuclear waste canister emplacement; structural analysis and design of nuclear waste package canister; robotic manipulation of the nuclear waste container; investigation of stress in a circular tunnel due to overburden & thermal loading of horizontally placed 21PWR multi-purpose canisters; investigation of faulted tunnel models by combined photoelasticity and finite element analysis; and transport phenomena in the near field.

  13. Analysis of near-field thermal and psychometric waste package environment using ventilation

    SciTech Connect

    Danko, G.; Buscheck, T.A.; Nitao, J.J.; Saterlie, S.

    1995-03-01

    The ultimate objective of the Civilian Radioactive Waste Management System (CRWMS) Program is to safely emplace and isolate the nations` spent nuclear fuel (SNF) and radioactive wastes in a geologic repository. Radioactive waste emplaced in a geologic repository will generate heat, increasing the temperature in the repository. The magnitude of this temperature increase depends upon (1) the heat source, i.e. the thermal loading of the repository, and (2) the geologic and engineered heat transport characteristics of the repository. Thermal management techniques currently under investigation include ventilation of the emplacement drifts during the preclosure period which could last as long as 100 years. Understanding the amount of heat and moisture removed from the emplacement drifts and near-field rock by ventilation, are important in determining performance of the engineered barrier system (EBS), as well as the corrosive environment of the waste packages, and the interaction of the EBS with the near-field host rock. Since radionuclide releases and repository system performance are significantly affected by the corrosion rate related to the psychometric environment, it is necessary to predict the amount of heat and moisture that are removed from the repository horizon using a realistic model for a wide range of thermal loading. This can be realized by coupling the hydrothermal model of the rock mass to a ventilation/climate model which includes the heat and moisture transport on the rock-air interface and the dilution of water vapor in the drift. This paper deals with the development of the coupled model concept, and determination of the boundary conditions for the calculations.

  14. Near-Field Hydrology Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    SciTech Connect

    PD Meyer; RJ Serne

    1999-12-21

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method for disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in new-surface, shallow land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford ILAW Performance Assessment (PA) Activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities and the consequent transport of dissolved contaminants in the pore water of the vadose zone. Pacific Northwest National Laboratory (PNNL) assists LMHC in its performance assessment activities. One of PNNL's tasks is to provide estimates of the physical, hydraulic, and transport properties of the materials comprising the disposal facilities and the disturbed region around them. These materials are referred to as the near-field materials. Their properties are expressed as parameters of constitutive models used in simulations of subsurface flow and transport. In addition to the best-estimate parameter values, information on uncertainty in the parameter values and estimates of the changes in parameter values over time are required to complete the PA. These parameter estimates and information are contained in this report, the Near-Field Hydrology Data Package.

  15. Iron chelating active packaging: Influence of competing ions and pH value on effectiveness of soluble and immobilized hydroxamate chelators.

    PubMed

    Ogiwara, Yoshiko; Roman, Maxine J; Decker, Eric A; Goddard, Julie M

    2016-04-01

    Many packaged foods utilize synthetic chelators (e.g. ethylenediaminetetraacetic acid, EDTA) to inhibit iron-promoted oxidation or microbial growth which would result in quality loss. To address consumer demands for all natural products, we have previously developed a non-migratory iron chelating active packaging material by covalent immobilization of polyhydroxamate and demonstrated its efficacy in delaying lipid oxidation. Herein, we demonstrate the ability of this hydroxamate-functionalized iron chelating active packaging to retain iron chelating capacity; even in the presence of competing ions common in food. Both immobilized and soluble hydroxamate chelators retained iron chelating capacity in the presence of calcium, magnesium, and sodium competing ions, although at pH 5.0 the presence of calcium reduced immobilized hydroxamate iron chelation. A strong correlation was found between colorimetric and mass spectral analysis of iron chelation by the chelating packaging material. Such chelating active packaging may support reducing additive use in product formulations, while retaining quality and shelf life.

  16. Evaluation and compilation of DOE waste package test data: Biannual report, February 1987--July 1987

    SciTech Connect

    Interrante, C.; Escalante, E.; Fraker, A.; Hall, D.; Harrison, S.; Liggett, W.; Linzer, M.; Ricker, R.; Ruspi, J.; Shull, R.

    1988-05-01

    The waste package is a proposed engineering barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon steels, stainless steels, and copper. The current level of understanding of several canister materials is questioned for the candidate repository in tuff. Three issues are addressed, the possibility of the stress-induced failure of Zircaloy, the possible corrosion of copper and copper alloys, and the lack of site-specific characterization data. Discussions are given on problems concerning localized corrosion and environmentally assisted cracking of AISI 1020 steel at elevated temperatures (150{degree}C). For the proposed salt site, the importance of the duration of corrosion tests and some of the conditions that may preclude prompt initiation of needed long-term testing are two issues that are discussed. 31 refs., 5 figs.

  17. Evaluation of the Corrosivity of Dust Deposited on Waste Packages at Yucca Mountain, Nevada

    SciTech Connect

    C. Bryan; R. Jarek; T. Wolery; D. Shields; M. Sutton; E. Hardin; D. Barr

    2005-03-01

    Potentially corrosive brines may form by deliquescence of salt minerals in dust deposited on the surface of waste packages at Yucca Mountain during operations and the pre-closure ventilation period. Evaluation of measured atmospheric and underground dust compositions by thermodynamic modeling and experimental studies of brine deliquescence indicates that brines are likely to form, but will be nitrate-rich and non-corrosive. Processes that modify the brines following deliquescence are beneficial with respect to corrosion. Acid degassing (HCI, HNO{sub 3}) will potentially dry out brines, but kinetic limitations will limit the effect to increasing the passivity of the brines by raising the pH and increasing the NO{sub 3}/CI ratio. Interactions with silicate minerals in the dust buffer brine pH to neutral values, and may also cause dryout. Predicted dust quantities and maximum deliquescent brine volumes on the waste package surface are small, and physical isolation of salt minerals in the dust may inhibit formation of eutectic brines, further decreasing deliquescence brine volumes. Should corrosive brines form, capillary forces will tend to retain brine in the dust. If brines do contact the WP surface, small droplet or film dimensions do not support development of diffusive gradients and separate anodic-cathodic regions necessary for initiation of localized corrosion. Finally, should localized corrosion initiate, corrosion product buildup will eventually stifle corrosion, by limiting oxygen access to the metal surface, by capillary retention of brine in corrosion product porosity, or by consumption of brine components, principally chloride. Thus, dust deliquescence is of low consequence with respect to repository performance.

  18. Reduced Pressure Electron Beam Welding Evaluation Activities on a Ni-Cr-Mo Alloy for Nuclear Waste Packages

    SciTech Connect

    Wong, F; Punshon, C; Dorsch, T; Fielding, P; Richard, D; Yang, N; Hill, M; DeWald, A; Rebak, R; Day, S; Wong, L; Torres, S; McGregor, M; Hackel, L; Chen, H-L; Rankin, J

    2003-09-11

    The current waste package design for the proposed repository at Yucca Mountain Nevada, USA, employs gas tungsten arc welding (GTAW) in fabricating the waste packages. While GTAW is widely used in industry for many applications, it requires multiple weld passes. By comparison, single-pass welding methods inherently use lower heat input than multi-pass welding methods which results in lower levels of weld distortion and also narrower regions of residual stresses at the weld TWI Ltd. has developed a Reduced Pressure Electron Beam (RPEB) welding process which allows EB welding in a reduced pressure environment ({le} 1 mbar). As it is a single-pass welding technique, use of RPEB welding could (1) achieve a comparable or better materials performance and (2) lead to potential cost savings in the waste package manufacturing as compared to GTAW. Results will be presented on the initial evaluation of the RPEB welding on a Ni-Cr-Mo alloy (a candidate alloy for the Yucca Mountain waste packages) in the areas of (a) design and manufacturing simplifications, (b) material performance and (c) weld reliability.

  19. Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

    SciTech Connect

    Delbert E. Day; Chandra S. Ray; Cheol-Woon Kim

    2004-12-28

    Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost.

  20. Lead-iron phosphate glass as a containment medium for the disposal of high-level nuclear wastes

    DOEpatents

    Boatner, L.A.; Sales, B.C.

    1984-04-11

    Disclosed are lead-iron phosphate glasses containing a high level of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

  1. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    SciTech Connect

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  2. Current trends for packaging transuranic waste at Los Alamos National Laboratory (LA-UR-07-4785)

    SciTech Connect

    Goyal, Kapil K.; Carson, Peter H.; Enriquez, Alejandro E.

    2007-07-01

    Transuranic (TRU) waste leaving the Plutonium Facility at Los Alamos National Laboratory (LANL) is packaged using LANL's waste acceptance criteria for onsite storage. Before shipment to the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, each payload container is subject to rigorous characterization to ensure compliance with WIPP waste acceptance criteria and Department of Transportation regulations. Techniques used for waste characterization include nondestructive examination by WIPP-certified real-time radiography (RTR) and nondestructive assay (NDA) of containers, as well as headspace gas sampling to ensure hydrogen and other flammable gases remain at safe levels during transport. These techniques are performed under a rigorous quality assurance program to confirm that results are accurate and reproducible. If containers are deemed problematic, corrective action is taken before shipment to WIPP. Currently this activity is possible only at the Laboratory's Waste Characterization, Reduction, and Repackaging Facility. To minimize additional waste requiring remediation, WIPP waste acceptance criteria must be applied at the point of waste generation. Additional criteria stem from limitations of RTR or NDA instruments or lack of appropriate sampling and analysis. This paper presents the changes that have been implemented at the Plutonium Facility and gives readers a preview of what LANL expects to accomplish to expeditiously certify and dispose of newly generated TRU waste. (authors)

  3. Assessment of collection schemes for packaging and other recyclable waste in European Union-28 Member States and capital cities.

    PubMed

    Seyring, Nicole; Dollhofer, Marie; Weißenbacher, Jakob; Bakas, Ioannis; McKinnon, David

    2016-09-01

    The Waste Framework Directive obliged European Union Member States to set up separate collection systems to promote high quality recycling for at least paper, metal, plastic and glass by 2015. As implementation of the requirement varies across European Union Member States, the European Commission contracted BiPRO GmbH/Copenhagen Resource Institute to assess the separate collection schemes in the 28 European Union Member States, focusing on capital cities and on metal, plastic, glass (with packaging as the main source), paper/cardboard and bio-waste. The study includes an assessment of the legal framework for, and the practical implementation of, collection systems in the European Union-28 Member States and an in depth-analysis of systems applied in all capital cities. It covers collection systems that collect one or more of the five waste streams separately from residual waste/mixed municipal waste at source (including strict separation, co-mingled systems, door-to-door, bring-point collection and civic amenity sites). A scoreboard including 13 indicators is elaborated in order to measure the performance of the systems with the capture rates as key indicators to identify best performers. Best performance are by the cities of Ljubljana, Helsinki and Tallinn, leading to the key conclusion that door-to-door collection, at least for paper and bio-waste, and the implementation of pay-as-you-throw schemes results in high capture and thus high recycling rates of packaging and other municipal waste.

  4. Isotope production potential at Sandia National Laboratories: Product, waste, packaging, and transportation

    SciTech Connect

    Trennel, A.J.

    1995-12-31

    The U.S. Congress directed the U.S. Department of Energy to establish a domestic source of molybdenum-99, an essential isotope used in nuclear medicine and radiopharmacology. An Environmental Impact Statement for production of {sup 99}Mo at one of four candidate sites is being prepared. As one of the candidate sites, Sandia National Laboratories is developing the Isotope Production Project. Using federally approved processes and procedures now owned by the U.S. Department of Energy, and existing facilities that would be modified to meet the production requirements, the Sandia National Laboratories` Isotope Project would manufacture up to 30 percent of the U.S. market, with the capacity to meet 100 percent of the domestic need if necessary. This paper provides a brief overview of the facility, equipment, and processes required to produce isotopes. Packaging and transportation issues affecting both product and waste are addressed, and the storage and disposal of the four low-level radioactive waste types generated by the production program are considered. Recommendations for future development are provided.

  5. Quantitative assessment of microbiological contributions to corrosion of candidate nuclear waste-package materials

    SciTech Connect

    Horn, J.; Jones, D.; Lian, T.; Martin, S.

    1998-10-30

    The U.S. Department of Energy is contributing to the design of a potential nuclear-waste repository at Yucca Mountain, Nevada. A system to predict the contribution of Yucca Mountain (YM) bacteria to overall corrosion rates of candidate waste-package (WP) materials was designed and implemented. DC linear polarization resistance techniques were applied to candidate material coupons that had been inoculated with a mixture of YM-derived bacteria with potentially corrosive activities or left sterile. Inoculated bacteria caused a 5- to 6-fold increase in corrosion rate of carbon steel C1020 (to approximately 7Ð8mm/yr) and an almost 100-fold increase in corrosion rate of Alloy 400 (to approximately 1mm/yr). Microbiologically influenced corrosion (MIC) rates on more resistant materials (CRMs: Alloy 625, Type 304 Stainless Steel, and Alloy C22) were on the order of hundredths of micrometers per year (mm/yr). Bulk chemical and surfacial end-point analyses of spent media and coupon surfaces showed preferential dissolution of nickel from Alloy 400 coupons and depletion of chromium from CRMs after incubation with YM bacteria. Scanning electron microscopy (SEM) also showed greater damage to the Alloy 400 surface than that indicated by electrochemical detection methods.

  6. Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants.

    PubMed

    Adrados, A; De Marco, I; Lopez-Urionabarrenechea, A; Caballero, B M; Laresgoiti, M F

    2013-01-01

    In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm(3) reactor, swept with 1 L min(-1) N(2), at 500°C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg(-1)). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO(2); their HHV is in the range of 18-46 MJ kg(-1). The amount of COCO(2) increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char.

  7. Geotechnical, Hydrogeologic and Vegetation Data Package for 200-UW-1 Waste Site Engineered Surface Barrier Design

    SciTech Connect

    Ward, Andy L.

    2007-11-26

    Fluor Hanford (FH) is designing and assessing the performance of engineered barriers for final closure of 200-UW-1 waste sites. Engineered barriers must minimize the intrusion and water, plants and animals into the underlying waste to provide protection for human health and the environment. The Pacific Northwest National Laboratory (PNNL) developed Subsurface Transport Over Multiple Phases (STOMP) simulator is being used to optimize the performance of candidate barriers. Simulating barrier performance involves computation of mass and energy transfer within a soil-atmosphere-vegetation continuum and requires a variety of input parameters, some of which are more readily available than others. Required input includes parameter values for the geotechnical, physical, hydraulic, and thermal properties of the materials comprising the barrier and the structural fill on which it will be constructed as well as parameters to allow simulation of plant effects. This report provides a data package of the required parameters as well as the technical basis, rationale and methodology used to obtain the parameter values.

  8. Welding Robot and Remote Handling System for the Yucca Mountain Waste Package Closure System

    SciTech Connect

    Barker, M.E.; Holt, T.E.; LaValle, D.R.; Pace, D.P.; Croft, K.M.; Shelton-Davis, C.V.

    2008-07-01

    In preparation for the license application and construction of a repository for housing the nation's spent nuclear fuel and high-level waste in Yucca Mountain, the Idaho National Laboratory (INL) has been charged with preparing a mock-up of a full-scale prototype system for sealing the waste packages (WP). Three critical pieces of the closure room include two PaR Systems TR4350 Telerobotic Manipulators and a PaR Systems XR100 Remote Handling System (RHS). The TR4350 Manipulators are 6-axis programmable robots that will be used to weld the WP lids and purge port cap as well as conduct nondestructive examinations. The XR100 Remote Handling System is a 4-axis programmable robot that will be used to transport the WP lids and process tools to the WP for operations and remove equipment for maintenance. The welding and RHS robots will be controlled using separate PaR 5/21 CIMROC Controllers capable of complex motion control tasks. A tele-operated PaR 4350 Manipulator will also be provided with the XR100 Remote Handling System. It will be used for maintenance and associated activities within the closure room. (authors)

  9. Hydrogen Concentration in the Inner-Most Container within a Pencil Tank Overpack Packaged in a Standard Waste Box Package

    SciTech Connect

    Marusich, Robert M.

    2012-01-25

    A set of steady state diffusion flow equations, for the hydrogen diffusion from one bag to the next bag (or one plastic waste container to another), within a set of nested waste bags (or nested waste containers), are developed and presented. The input data is then presented and justified. Inputting the data for each volume and solving these equations yields the steady state hydrogen concentration in each volume. The input data (permeability of the bag surface and closure, dimensions and hydrogen generation rate) and equations are analyzed to obtain the hydrogen concentrations in the innermost container for a set of containers which are analyzed for the TRUCON code for the general waste containers and the TRUCON code for the Pencil Tank Overpacks (PTO) in a Standard Waste Box (SWB).

  10. Property and process correlations for iron-enriched basalt waste forms

    SciTech Connect

    Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1993-02-01

    Correlations of thermodynamic properties and process parameters of high-temperature slag for a range of compositions of iron-enriched basalt are presented. The quantification of the properties of this complex mixture can assist in the design and monitoring of high-temperature melting systems for the treatment of radioactive and hazardous wastes at the Idaho National Engineering Laboratory. The buried and stored wastes at the INEL Radioactive Waste Management Complex have a similar composition to iron-enriched basalt after oxidation of organics. The properties correlated are the viscosity, electrical conductivity, refractory corrosion, and recrystallization temperature. The correlations are expressed as a function of input waste-soil mixture composition, alkali concentration, and slag temperature. An application to determine the effect of alkali flux on slag temperature, leach rate, and volume reduction is presented. Though the correlations are for mixtures of soil and waste with average transuranic-contaminated waste compositions, it appears that good approximations for other waste streams and glass-ceramic waste forms can be obtained because of similarities in composition.

  11. Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants

    SciTech Connect

    Adrados, A.

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Study of the influence of materials in the pyrolysis of real plastic waste samples. Black-Right-Pointing-Pointer Inorganic compounds remain unaltered. Black-Right-Pointing-Pointer Cellulosic components give rise to an increase in char formation. Black-Right-Pointing-Pointer Cellulosic components promote the production of aqueous phase. Black-Right-Pointing-Pointer Cellulosic components increase CO and CO{sub 2} contents in the gases. - Abstract: In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm{sup 3} reactor, swept with 1 L min{sup -1} N{sub 2}, at 500 Degree-Sign C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg{sup -1}). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO{sub 2}; their HHV is in the range of 18-46 MJ kg{sup -1}. The amount of CO-CO{sub 2} increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char.

  12. Nuclear waste transportation package testing: A review of selected programs in the United States and abroad

    SciTech Connect

    Snedeker, D F

    1990-12-01

    This report provides an overview of some recent nuclear waste transportation package development programs. This information is intended to aid the State of Nevada in its review of US Department of Energy (DOE) nuclear waste transportation programs. This report addresses cask testing programs in the United Kingdom and selected 1/4 and full scale testing in the US. Facilities that can provide cask testing services, both in the US and to a limited extent abroad, are identified. The costs for different type test programs are identified as a means to estimate costs for future test programs. Not addressed is the public impact such testing might have in providing an increased sense of safety or confidence. The British test program was apparently quite successful in demonstrating safety to the public at the time. There is no US test effort that is similar in scope for direct comparison. Also addressed are lessons learned from testing programs and areas that may merit possible future integrated examination. Areas that may require further examination are both technical and institutional. This report provides information which, when combined with other sources of information will enable the State of Nevada to assess the following areas: feasibility of full scale testing; costs of full scale tests; potential benefits of testing; limits that full scale testing impose; and disadvantages of emphasis on testing vs analytical solutions. This assessment will then allow the state to comment on DOE Office of Civilian Radioactive Waste Management (OCRWM) plans for the development and licensing of new shipping cask designs. These plans currently expect contractors to perform engineering testing for materials development, quarter scale model testing to validate analytical assessments and full scale prototype testing of operational features. DOE currently plans no full scale or extra-regulatory destructive testing to aid in cask licensing. 1 tab.

  13. Reducing acid leaching of manganiferous ore: effect of the iron removal operation on solid waste disposal.

    PubMed

    De Michelis, Ida; Ferella, Francesco; Beolchini, Francesca; Vegliò, Francesco

    2009-01-01

    The process of reducing acid leaching of manganiferous ore is aimed at the extraction of manganese from low grade manganese ores. This work is focused on the iron removal operation. The following items have been considered in order to investigate the effect of the main operating conditions on solid waste disposal and on the process costs: (i) type and quantity of the base agent used for iron precipitation, (ii) effective need of leaching waste separation prior to the iron removal operation, (iii) presence of a second leaching stage with the roasted ore, which might also act as a preliminary iron removal step, and (iv) effect of tailings washing on the solid waste classification. Different base compounds have been tested, including CaO, CaCO3, NaOH, and Na2CO3. The latter gave the best results concerning both the precipitation process kinetics and the reagent consumption. The filtration of the liquor leach prior to iron removal was not necessary, implying significant savings in capital costs. A reduction of chemical consumption and an increase of manganese concentration in the solution were obtained by introducing secondary leaching tests with the previously roasted ore; this additional step was introduced without a significant decrease of global manganese extraction yield. Finally, toxicity characteristic leaching procedure (TCLP) tests carried out on the leaching solid waste showed: (i) a reduction of arsenic mobility in the presence of iron precipitates, and (ii) the need for a washing step in order to produce a waste that is classifiable as not dangerous, taking into consideration the existing Environmental National Laws.

  14. Packaging waste prevention in the distribution of fruit and vegetables: An assessment based on the life cycle perspective.

    PubMed

    Tua, Camilla; Nessi, Simone; Rigamonti, Lucia; Dolci, Giovanni; Grosso, Mario

    2017-02-01

    In recent years, alternative food supply chains based on short distance production and delivery have been promoted as being more environmentally friendly than those applied by the traditional retailing system. An example is the supply of seasonal and possibly locally grown fruit and vegetables directly to customers inside a returnable crate (the so-called 'box scheme'). In addition to other claimed environmental and economic advantages, the box scheme is often listed among the packaging waste prevention measures. To check whether such a claim is soundly based, a life cycle assessment was carried out to verify the real environmental effectiveness of the box scheme in comparison to the Italian traditional distribution. The study focused on two reference products, carrots and apples, which are available in the crate all year round. An experience of a box scheme carried out in Italy was compared with some traditional scenarios where the product is distributed loose or packaged at the large-scale retail trade. The packaging waste generation, 13 impact indicators on environment and human health and energy consumptions were calculated. Results show that the analysed experience of the box scheme, as currently managed, cannot be considered a packaging waste prevention measure when compared with the traditional distribution of fruit and vegetables. The weaknesses of the alternative system were identified and some recommendations were given to improve its environmental performance.

  15. Summary of INEL research on the iron-enriched basalt waste form

    SciTech Connect

    Reimann, G.A.; Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1992-01-01

    This report summarizes the knowledge base on the iron-enriched basalt (IEB) waste form developed at the Idaho National Engineering Laboratory (INEL) during 1979--1982. The results presented discuss the applicability of IEB in converting retrieved transuranic (TRU) waste from INEL's Radioactive Waste Management Complex (RWMC) into a vitreous/ceramic (glassy/rock) stable waste form suitable for permanent disposal in an appropriate repository, such as the Waste Isolation Pilot Plant (WIPP) in New Mexico. Borosilicate glass (BSG), the approved high-level waste form, appears unsuited for this application. Melting the average waste-soil mix from the RWMC produces the IEB composition and attempting to convert IEB to the BSG composition would require additions of substantial B{sub 2}0{sub 3}, Na, and SiO{sub 2} (glass frit). IEB requires processing temperatures of 1400 to 1600{degrees}C, depending upon the waste composition. Production of the IEB waste form, using Joule heated melters, has proved difficult in the past because of electrode and refractory corrosion problems associated with the high temperature melts. Higher temperature electric melters (arc and plasma) are available to produce this final waste form. Past research focused on extensive slag property measurements, waste form leachability tests, mechanical, composition, and microstructure evaluations, as well as a host of experiments to improve production of the waste form. Past INEL studies indicated that the IEB glass-ceramic is a material that will accommodate and stabilize a wide range of heterogeneous waste materials, including long lived radionuclides and scrap metals, while maintaining a superior level of chemical and physical performance characteristics. Controlled cooling of the molten IEB and subsequent heat treatment will produce a glass-ceramic waste form with superior leach resistance.

  16. Summary of INEL research on the iron-enriched basalt waste form

    SciTech Connect

    Reimann, G.A.; Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1992-01-01

    This report summarizes the knowledge base on the iron-enriched basalt (IEB) waste form developed at the Idaho National Engineering Laboratory (INEL) during 1979--1982. The results presented discuss the applicability of IEB in converting retrieved transuranic (TRU) waste from INEL`s Radioactive Waste Management Complex (RWMC) into a vitreous/ceramic (glassy/rock) stable waste form suitable for permanent disposal in an appropriate repository, such as the Waste Isolation Pilot Plant (WIPP) in New Mexico. Borosilicate glass (BSG), the approved high-level waste form, appears unsuited for this application. Melting the average waste-soil mix from the RWMC produces the IEB composition and attempting to convert IEB to the BSG composition would require additions of substantial B{sub 2}0{sub 3}, Na, and SiO{sub 2} (glass frit). IEB requires processing temperatures of 1400 to 1600{degrees}C, depending upon the waste composition. Production of the IEB waste form, using Joule heated melters, has proved difficult in the past because of electrode and refractory corrosion problems associated with the high temperature melts. Higher temperature electric melters (arc and plasma) are available to produce this final waste form. Past research focused on extensive slag property measurements, waste form leachability tests, mechanical, composition, and microstructure evaluations, as well as a host of experiments to improve production of the waste form. Past INEL studies indicated that the IEB glass-ceramic is a material that will accommodate and stabilize a wide range of heterogeneous waste materials, including long lived radionuclides and scrap metals, while maintaining a superior level of chemical and physical performance characteristics. Controlled cooling of the molten IEB and subsequent heat treatment will produce a glass-ceramic waste form with superior leach resistance.

  17. Study of the possibility of iron extraction from waste coal by magnetizing roasting

    NASA Astrophysics Data System (ADS)

    Volynkina, E. P.; Zorya, V. N.; Korovushkin, V. V.

    2016-09-01

    The efficiency of magnetizing roasting of waste coal in converting iron, contained in the mineral part, from weak-magnetic and non-magnetic minerals (siderite, pyrite, marcasite) into high-magnetic ones (magnetite, maghemite) is shown. Products of magnetizing roasting are used as raw material for the subsequent magnetic separation.

  18. MINERALOGY AND CHARACTERIZATION OF ARSENIC, IRON, AND LEAD IN A MINE WASTE-DERIVED FERTILIZER

    EPA Science Inventory

    The solid-state speciation of arsenic (As), iron (Fe), and lead (Pb) was studied in the mine waste-derived fertilizer Ironite using X-ray absorption spectroscopy, Mössbauer spectroscopy, and aging studies. Arsenic was primarily associated with ferrihydrite (60-70%) with the rema...

  19. Recharge Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    SciTech Connect

    MJ Fayer; EM Murphy; JL Downs; FO Khan; CW Lindenmeier; BN Bjornstad

    2000-01-18

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is known as the Hanford ILAW Performance Assessment (PA) Activity, hereafter called the ILAW PA project. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require predictions of contaminant migration from the facility. To make such predictions will require estimates of the fluxes of water moving through the sediments within the vadose zone around and beneath the disposal facility. These fluxes, loosely called recharge rates, are the primary mechanism for transporting contaminants to the groundwater. Pacific Northwest National Laboratory (PNNL) assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of recharge rates for current conditions and long-term scenarios involving the shallow-land disposal of ILAW. Specifically, recharge estimates are needed for a filly functional surface cover; the cover sideslope, and the immediately surrounding terrain. In addition, recharge estimates are needed for degraded cover conditions. The temporal scope of the analysis is 10,000 years, but could be longer if some contaminant peaks occur after 10,000 years. The elements of this report compose the Recharge Data Package, which provides estimates of recharge rates for the scenarios being considered in the 2001 PA. Table S.1 identifies the surface features and

  20. [Methodological approaches to the development of environmentally benign technology for the use of solid waste in iron metallurgy].

    PubMed

    Pugin, K G; Vaĭsman, Ia I

    2013-01-01

    On the basis of the life cycle of materials, containing wastes of iron and steel industry, new methodological approaches to the assessment of technologies of the secondary use of wastes are developed A complex criteria for selection of the technology for the use of resource potential of solid waste of iron and steel industry are developed with taking into account environmental, technological and economic indices. The technology of the use of wastes of ferrovanadium industry as bulk solid materials at the solid waste landfill is shown.

  1. Geochemical data package for the Hanford immobilized low-activity tank waste performance assessment (ILAW PA)

    SciTech Connect

    DI Kaplan; RJ Serne

    2000-02-24

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as low-activity waste is to vitrify the liquid/slurry and place the solid product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (PA) activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities, and the consequent transport of dissolved contaminants in the porewater of the vadose zone. Pacific Northwest National Laboratory assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of the geochemical properties of the materials comprising the disposal facility, the disturbed region around the facility, and the physically undisturbed sediments below the facility (including the vadose zone sediments and the aquifer sediments in the upper unconfined aquifer). The geochemical properties are expressed as parameters that quantify the adsorption of contaminants and the solubility constraints that might apply for those contaminants that may exceed solubility constraints. The common parameters used to quantify adsorption and solubility are the distribution coefficient (K{sub d}) and the thermodynamic solubility product (K{sub sp}), respectively. In this data package, the authors approximate the solubility of contaminants using a more simplified construct

  2. Tritium Packages and 17th RH Canister Categories of Transuranic Waste Stored Below Ground within Area G

    SciTech Connect

    Hargis, Kenneth Marshall

    2015-03-01

    A large wildfire called the Las Conchas Fire burned large areas near Los Alamos National Laboratory (LANL) in 2011 and heightened public concern and news media attention over transuranic (TRU) waste stored at LANL’s Technical Area 54 (TA-54) Area G waste management facility. The removal of TRU waste from Area G had been placed at a lower priority in budget decisions for environmental cleanup at LANL because TRU waste removal is not included in the March 2005 Compliance Order on Consent (Reference 1) that is the primary regulatory driver for environmental cleanup at LANL. The Consent Order is a settlement agreement between LANL and the New Mexico Environment Department (NMED) that contains specific requirements and schedules for cleaning up historical contamination at the LANL site. After the Las Conchas Fire, discussions were held by the U.S. Department of Energy (DOE) with the NMED on accelerating TRU waste removal from LANL and disposing it at the Waste Isolation Pilot Plant (WIPP). This report summarizes available information on the origin, configuration, and composition of the waste containers within the Tritium Packages and 17th RH Canister categories; their physical and radiological characteristics; the results of the radioassays; and potential issues in retrieval and processing of the waste containers.

  3. Upgrading of recycled plastics obtained from flexible packaging waste by adding nanosilicates

    NASA Astrophysics Data System (ADS)

    Garofalo, E.; Claro, M.; Scarfato, P.; Di Maio, L.; Incarnato, L.

    2015-12-01

    Currently, the growing consumption of polymer products creates large quantities of waste materials resulting in public concern in the environment and people life. The efficient treatment of polymer wastes is still a difficult challenge and the recycling process represents the best way to manage them. Recently, many researchers have tried to develop nanotechnology for polymer recycling. The products prepared through the addition of nanoparticles to post-used plastics could offer the combination of improved properties, low weight, easy of processing and low cost which is not easily and concurrently found by other methods of plastic recycling. In this study materials, obtained by the separation and mechanical recycling of post-consumer packaging films of small size (

  4. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    SciTech Connect

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr.; Gdowski, G.E.

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

  5. Evaluation and compilation of DOE [Department of Energy] waste package test data; Biannual report, February 1988--July 1988

    SciTech Connect

    Interrante, C.; Escalante, E.; Fraker, A.; Plante, E.

    1989-10-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six month period February 1988 through July 1988. Activities for the DOE Materials Characterization Center are reviewed for the period January 1988 through June 1988. A summary is given of the Yucca Mountain, Nevada disposal site activities. Short discussions relating to the reviewed publications are given and complete reviews and evaluations are included. 20 refs., 1 fig., 1 tab.

  6. Evaluation and compilation of DOE waste package test data; Volume 8: Biannual report, August 1989--January 1990

    SciTech Connect

    Interrante, C.G.; Fraker, A.C.; Escalante, E.

    1993-06-01

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

  7. Scoping corrosion tests on candidate waste package basket materials for the Yucca Mountain project

    SciTech Connect

    Konynenburg, R.A. van; Curtis, P.G.; Summers, T.S.E.

    1998-03-01

    A scoping corrosion test was performed on candidate waste package basket materials. The corrosion medium was a pH-buffered solution of chemical species expected to be produced by radiolysis. The test was conducted at 90 C for 96 hours. Samples included aluminum-, copper-, stainless steel- and zirconium-based metallic materials and several ceramics, incorporating neutron-absorbing elements. Sample weight losses and solution chemical changes were measured. Both corrosion of the host materials and dissolution of the neutron-absorbing elements were studied. The ceramics and the zirconium-based materials underwent only minor corrosion. The stainless steel-based materials performed well except for a welded sample. The aluminum- and copper-based materials exhibited the highest corrosion rates. Boron dissolution depends on its chemical form. Boron oxide and many metal borides dissolve readily in acidic solutions while high-chromium borides and boron carbide, though thermodynamically unstable, exhibit little dissolution in short times. The results of solution chemical analyses were consistent with this. Gadolinium did not dissolve significantly from monazite, and hafnium showed little dissolution from a variety of host materials, in keeping with its low solubility.

  8. Scoping corrosion tests on candidate waste package basket materials for the Yucca Mountain Project

    SciTech Connect

    Van Konynenburg, R.A.; Curits, P.C.; Summers, T.S.E.

    1998-03-01

    A scoping corrosion test was performed on candidate waste package basket materials. The corrosion medium was a pH-buffered solution of chemical species expected to be produced by radiolysis. The test was conducted at 90{degrees}C for 96 hours. Samples included aluminum-, copper-, stainless steel-, and zirconium-based metallic materials and several ceramics, incorporating neutron-absorbing elements. Sample weight losses and solution chemical changes were measured. Both corrosion of the host materials and dissolution of the neutron- absorbing elements were studied. The ceramics and the zirconium-based materials underwent only minor corrosion. the stainless steel-based materials performed well except for a welded sample. The aluminum- and copper-based materials exhibited the highest corrosion rates. Boron dissolution depends on it chemical form. Boron oxide and many metal borides dissolve readily in acidic solutions while high- chromium borides and boron carbide, though thermodynamically unstable, exhibit little dissolution in short times. the results of solution chemical analyses were consistent with this. Gadolinium did not dissolve significantly from monazite, and hafnium showed little dissolution from a variety of host materials, in keeping with its low solubility.

  9. Feasibility study of the separation of chlorinated films from plastic packaging wastes

    SciTech Connect

    Reddy, Mallampati Srinivasa; Yamaguchi, Takefumi; Okuda, Tetsuji; Tsai, Tsung-Yueh; Nakai, Satoshi; Nishijima, Wataru; Okada, Mitsumasa

    2010-04-15

    This study describes the possible separation of chlorinated plastic films (PVC and PVDC) from other heavy plastic packaging waste (PPW) by selective twist formation and gravity separation. Twists formation was mechanically induced in chlorinated plastic films, whereas twist formation did not occur in PS and PET films. After twist formation, all the films had the apparent density of less than 1.0 g/cm{sup 3} and floated in water even though the true density was more than 1.0 g/cm{sup 3}. However, the apparent density of the PS and the PET films increased with agitation to more than 1.0 g/cm{sup 3}, whereas that of chlorinated plastic films was kept less than 1.0 g/cm{sup 3}. The main reason would be the air being held inside the chlorinated plastic films which was difficult to be removed by agitation. Simple gravity separation after twist formation was applied for artificial film with 10 wt.% of the chlorinated films and real PPW films with 9 wt.% of the chlorinated films. About 76 wt.% of the artificial PPW films and 75 wt.% of real PPW films after the removal of PP and PE were recovered as settling fraction with 4.7 wt.% and 3.0 wt.% of chlorinated plastic films, respectively. These results indicate that simple gravity separation process after twist formation can be used to reduce the chlorinated plastic concentration from mixed heavy PPW films.

  10. Feasibility study of the separation of chlorinated films from plastic packaging wastes.

    PubMed

    Reddy, Mallampati Srinivasa; Yamaguchi, Takefumi; Okuda, Tetsuji; Tsai, Tsung-Yueh; Nakai, Satoshi; Nishijima, Wataru; Okada, Mitsumasa

    2010-04-01

    This study describes the possible separation of chlorinated plastic films (PVC and PVDC) from other heavy plastic packaging waste (PPW) by selective twist formation and gravity separation. Twists formation was mechanically induced in chlorinated plastic films, whereas twist formation did not occur in PS and PET films. After twist formation, all the films had the apparent density of less than 1.0g/cm(3) and floated in water even though the true density was more than 1.0g/cm(3). However, the apparent density of the PS and the PET films increased with agitation to more than 1.0g/cm(3), whereas that of chlorinated plastic films was kept less than 1.0g/cm(3). The main reason would be the air being held inside the chlorinated plastic films which was difficult to be removed by agitation. Simple gravity separation after twist formation was applied for artificial film with 10wt.% of the chlorinated films and real PPW films with 9wt.% of the chlorinated films. About 76wt.% of the artificial PPW films and 75wt.% of real PPW films after the removal of PP and PE were recovered as settling fraction with 4.7wt.% and 3.0wt.% of chlorinated plastic films, respectively. These results indicate that simple gravity separation process after twist formation can be used to reduce the chlorinated plastic concentration from mixed heavy PPW films.

  11. Feasibility assessment of copper-base waste package container materials in a tuff repository

    SciTech Connect

    Acton, C.F.; McCright, R.D.

    1986-09-30

    This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized.

  12. Hydrogen Concentration in the Inner-Most Container within a Pencil Tank Overpack Packaged in a Standard Waste Box Package

    SciTech Connect

    Marusich, Robert M.

    2013-08-15

    The purpose of this report is to evaluate hydrogen generation within Pencil Tank Overpacks (PTO) in a Standard Waste Box (SWB), to establish plutonium (Pu) limits for PTOs based on hydrogen concentration in the inner-most container and to establish required configurations or validate existing or proposed configurations for PTOs. The methodology and requirements are provided in this report.

  13. Reaction of Inconel 690 and 693 in Iron Phosphate Melts: Alternative Glasses for Waste Vitrification

    SciTech Connect

    Day, Delbert E. Kim, Cheol-Woon

    2005-09-13

    The corrosion resistance of candidate materials used for the electrodes (Inconel 690 & 693) and the melt contact refractory (Monofrax K-3) in a Joule Heated Melter (JHM) has been investigated at the University of Missouri-Rolla (UMR) during the period from June 1, 2004 to August 31, 2005. This work was supported by the U.S. Department of Energy (DOE) Office of Biological and Environmental Research (DE-FG02-04ER63831). The unusual properties and characteristics of iron phosphate glasses, as viewed from the standpoint of alternative glasses for vitrifying nuclear and hazardous wastes which contain components that make them poorly suited for vitrification in borosilicate glass, were recently discovered at UMR. The expanding national and international interest in iron phosphate glasses for waste vitrification stems from their rapid melting and chemical homogenization which results in higher furnace output, their high waste loading that varies from 32 wt% up to 75 wt% for the Hanford LAW and HLW, respectively, and the outstanding chemical durability of the iron phosphate wasteforms which meets all present DOE requirements (PCT and VHT). The higher waste loading in iron phosphate glasses, compared to the baseline borosilicate glass, can reduce the time and cost of vitrification considerably since a much smaller mass of glass will be produced, for example, about 43% less glass when the LAW at Hanford is vitrified in an iron phosphate glass according to PNNL estimates. In view of the promising performance of iron phosphate glasses, information is needed for how to best melt these glasses on the scale needed for practical use. Melting iron phosphate glasses in a JHM is considered the preferred method at this time because its design could be nearly identical to the JHM now used to melt borosilicate glasses at the Defense Waste Processing Facility (DWPF), Westinghouse Savannah River Co. Therefore, it is important to have information for the corrosion of candidate electrode

  14. FINITE-ELEMENT ANALYSIS OF ROCK FALL ON UNCANISTERED FUEL WASTE PACKAGE DESIGNS (SCPB: N/A)

    SciTech Connect

    Z. Ceylan

    1996-10-18

    The objective of this analysis is to explore the Uncanistered Fuel (UCF) Tube Design waste package (WP) resistance to rock falls. This analysis will also be used to determine the size of rock that can strike the WP without causing failure in the containment barriers from a height based on the starter tunnel dimensions. The purpose of this analysis is to document the models and methods used in the calculations.

  15. Iron-phosphate ceramics for solidification of mixed low-level waste

    DOEpatents

    Aloy, Albert S.; Kovarskaya, Elena N.; Koltsova, Tatiana I.; Macheret, Yevgeny; Medvedev, Pavel G.; Todd, Terry

    2000-01-01

    A method of immobilizing mixed low-level waste is provided which uses low cost materials and has a relatively long hardening period. The method includes: forming a mixture of iron oxide powders having ratios, in mass %, of FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 equal to 25-40:40-10:35-50, or weighing a definite amount of magnetite powder. Metallurgical cinder can also be used as the source of iron oxides. A solution of the orthophosphoric acid, or a solution of the orthophosphoric acid and ferric oxide, is formed and a powder phase of low-level waste and the mixture of iron oxide powders or cinder (or magnetite powder) is also formed. The acid solution is mixed with the powder phase to form a slurry with the ratio of components (mass %) of waste:iron oxide powders or magnetite:acid solution=30-60:15-10:55-30. The slurry is blended to form a homogeneous mixture which is cured at room temperature to form the final product.

  16. FY 1985 status report on feasibility assessment of copper-base waste package container materials in a tuff repository

    SciTech Connect

    McCright, R.D.

    1985-09-30

    This report discusses progress made during the first year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. The expected corrosion and oxidation performances of oxygen-free copper, aluminum bronze, and 70% copper-30% nickel are presented; a test plan for determining whether copper or one of the alloys can meet the containment requirements is outlined. Some preliminary corrosion test data are presented and discussed. Fabrication and joining techniques for forming waste package containers are descibed. Preliminary test data and analyses indicate that copper and copper-base alloys have several attractive features as waste package container materials, but additional work is needed before definitive conclusions can be made on the feasibility of using copper or a copper-base alloy for containers. Plans for work to be undertaken in the second year are indicated.

  17. Iron-phosphate-based chemically bonded phosphate ceramics for mixed waste stabilization

    SciTech Connect

    Wagh, A.S.; Jeong, S.Y.; Singh, D.

    1997-01-01

    In an effort to develop chemically bonded phosphate ceramics for mixed waste stabilization, a collaborative project to develop iron-phosphate based ceramics has been initiated between Argonne National Laboratory and the V. G. Khlopin Radium Institute in St. Petersburg, Russia. The starter powders are oxides of iron that are generated as inexpensive byproduct materials in the iron and steel industry. They contain iron oxides as a mixture of magnetite (Fe{sub 3}O{sub 4}) and haematite (Fe{sub 2}O{sub 3}). In this initial phase of this project, both of these compounds were investigated independently. Each was reacted with phosphoric acid solution to form iron phosphate ceramics. In the case of magnetite, the reaction was rapid. Adding ash as the waste component containing hazardous contaminants resulted in a dense and hard ceramic rich in glassy phase. On the other hand, the reaction of phosphoric acid solution with a mixture of haematite and ash waste contaminated with cesium and americium was too slow. Samples had to be molded under pressure. They were cured for 2-3 weeks and then hardened by heating at 350{degrees}C for 3 h. The resulting ceramics in both cases were subjected to physical tests for measurement of density, open porosity, compression strength, phase analyses using X-ray diffraction and differential thermal analysis, and leaching tests using toxicity characteristic leaching procedure (TCLP) and ANS 16.1 with 7 days of leaching. Using the preliminary information obtained from these tests, we evaluated these materials for stabilization of Department of Energy`s mixed waste streams.

  18. The influence of iron concentration on biohydrogen production from organic waste via anaerobic fermentation.

    PubMed

    Boni, M R; Sbaffoni, S; Tuccinardi, L

    2014-01-01

    Different micronutrients are essential for bacterial fermentative metabolism. In particular, some metallic ions, like iron, are able to affect the biological H₂production. In this study, batch tests were carried out in stirred reactors to investigate the effects of Fe²⁺ concentration on fermentative H₂production from two different organic fractions of waste: source-separated organic waste (OW) from a composting plant including organic fraction of municipal solid waste and food waste (FW) from a refectory. Iron supplementation at 1000 mg/L caused twofold increment in the cumulative H₂production from OW (922 mL) compared with the control (without iron doping). The highest H₂production (1736 mL) from FW occurred when Fe²⁺ concentration was equal to 50 mg/L. In addition, the process production from OW was modelled through the modified Gompertz equation. For FW, a translated Gompertz equation was used by the authors, because the initial lag-time for H₂production from FW was almost negligible.

  19. Development of Iron-Chelating Poly(ethylene terephthalate) Packaging for Inhibiting Lipid Oxidation in Oil-in-Water Emulsions.

    PubMed

    Johnson, David R; Tian, Fang; Roman, Maxine J; Decker, Eric A; Goddard, Julie M

    2015-05-27

    Foods such as bulk oils, salad dressings, and nutritionally fortified beverages that are susceptible to oxidative degradation are often packaged in poly(ethylene terephthalate) (PET) bottles with metal chelators added to the food to maintain product quality. In the present work, a metal-chelating active packaging material is designed and characterized, in which poly(hydroxamic acid) (PHA) metal-chelating moieties were grafted from the surface of PET. Biomimetic PHA groups were grafted in a two-step UV-initiated process without the use of a photoinitiator. Surface characterization of the films by attenuated total reflective Fourier transform infrared spectroscopy (ATR-FTIR) and scanning electron microscopy (SEM) suggested successful grafting and conversion of poly(hydroxyethyl acrylate) (PHEA) to PHA chelating moieties from the surface of PET. Colorimetric (ferrozine) and inductively coupled plasma mass spectroscopy (ICP-MS) assays demonstrated the ability of PET-g-PHA to chelate iron in a low-pH (3.0) environment containing a competitive metal chelator (citric acid). Lipid oxidation studies demonstrated the antioxidant activity of PET-g-PHA films in inhibiting iron-promoted oxidation in an acidified oil-in-water (O/W) emulsion model system (pH 3.0). Particle size and ζ-potential analysis indicated that the addition of PET-g-PHA films did not affect the physical stability of the emulsion system. This work suggests that biomimetic chelating moieties can be grafted from PET and effectively inhibit iron-promoted degradation reactions, enabling removal of metal-chelating additives from product formulations.

  20. Operating Experience and Lessons Learned in the Use of Soft-Sided Packaging for Transportation and Disposal of Low Activity Radioactive Waste

    SciTech Connect

    Kapoor, A.; Gordon, S.; Goldston, W.

    2013-07-08

    This paper describes the operating experience and lessons learned at U.S. Department of Energy (DOE) sites as a result of an evaluation of potential trailer contamination and soft-sided packaging integrity issues related to the disposal of low-level and mixed low-level (LLW/MLLW) radioactive waste shipments. Nearly 4.3 million cubic meters of LLW/MLLW will have been generated and disposed of during fiscal year (FY) 2010 to FY 2015—either at commercial disposal sites or disposal sites owned by DOE. The LLW/MLLW is packaged in several different types of regulatory compliant packaging and transported via highway or rail to disposal sites safely and efficiently in accordance with federal, state, and local regulations and DOE orders. In 1999, DOE supported the development of LLW containers that are more volumetrically efficient, more cost effective, and easier to use as compared to metal or wooden containers that existed at that time. The DOE Idaho National Engineering and Environmental Laboratory (INEEL), working in conjunction with the plastic industry, tested several types of soft-sided waste packaging systems that meet U.S. Department of Transportation requirements for transport of low specific activity and surface contaminated objects. Since then, soft-sided packaging of various capacities have been used successfully by the decontamination and decommissioning (D&D) projects to package, transport, and dispose D&D wastes throughout the DOE complex. The joint team of experts assembled by the Energy Facility Contractors Group from DOE waste generating sites, DOE and commercial waste disposal facilities, and soft-sided packaging suppliers conducted the review of soft-sided packaging operations and transportation of these packages to the disposal sites. As a result of this evaluation, the team developed several recommendations and best practices to prevent or minimize the recurrences of equipment contamination issues and proper use of soft-sided packaging for transport

  1. Utilization of food industry wastes for the production of zero-valent iron nanoparticles.

    PubMed

    Machado, S; Grosso, J P; Nouws, H P A; Albergaria, J T; Delerue-Matos, C

    2014-10-15

    The proper disposal of the several types of wastes produced in industrial activities increases production costs. As a consequence, it is common to develop strategies to reuse these wastes in the same process and in different processes or to transform them for use in other processes. This work combines the needs for new synthesis methods of nanomaterials and the reduction of production cost using wastes from citrine juice (orange, lime, lemon and mandarin) to produce a new added value product, green zero-valent iron nanoparticles that can be used in several applications, including environmental remediation. The results indicate that extracts of the tested fruit wastes (peel, albedo and pulp fractions) can be used to produce zero-valent iron nanoparticles (nZVIs). This shows that these wastes can be an added value product. The resulting nZVIs had sizes ranging from 3 up to 300 nm and distinct reactivities (pulp>peel>albedo extracts). All the studied nanoparticles did not present a significant agglomeration/settling tendency when compared to similar nanoparticles, which indicates that they remain in suspension and retain their reactivity.

  2. Removal of trichloroethylene by zerovalent iron/activated carbon derived from agricultural wastes.

    PubMed

    Su, Yuh-fan; Cheng, Yu-ling; Shih, Yang-hsin

    2013-11-15

    Activated carbon (AC) and zerovalent iron (ZVI) have been widely used in the adsorption and dehalogenation process, respectively, for the removal of organic compounds in environmental treatments. This study aims to prepare ZVI/AC derived from an agricultural waste, coir pith, through simple one-step pyrolysis. The effect of activation temperature and time on the surface area, iron content, and zerovalent iron ratio of ZVI/AC was systemically investigated. The results indicated that the activation of AC by FeSO4 significantly increased surface area of AC and distributed elemental iron over the AC. The X-ray diffraction (XRD), electron spectroscopy for chemical analysis (ESCA), and X-ray absorption near edge structure (XANES) spectra of ZVI/AC revealed that zerovalent iron was present. As compared to AC without FeSO4 activation, ZVI/AC increased the trichloroethylene removal rate constant by 7 times. The dechlorination ability of ZVI/AC was dominated by the zerovalent iron content. We have shown that lab-made ZVI/AC from coir pith can effectively adsorb and dehalogenate the chlorinated compounds in water.

  3. Conceptual designs for waste packages for horizontal or vertical emplacement in a repository in salt for reference in the site characterization plan

    SciTech Connect

    Not Available

    1987-06-01

    This report includes the options of horizontal and vertical emplacement, the addition of a phased repository, an additional waste form (intact spent fuel), revised geotechnical data appropriate for the Deaf Smith County site, new corrosion data for the container, and new repository design data. The waste package consists of waste form and canister within a thick-walled, low-carbon steel container surrounded by packing. The container is a hollow cylinder with a flat head welded to each end. The design concepts for the waste container or vertical and horizontal emplacement are identical. This report discusses the results of analyses of aspects of the reference waste package concept needing changes because of new data and information believed applicable to the Deaf Smith County site. Included are waste package conceptual designs or (1) the reference defense high-level waste form from the Savannah River Plant; (2) intact spent fuel with our pressurized-water-reactor or nine boiling-water-reactor assemblies per package for emplacement during Phase 1 of repository operation; and (3) spent fuel which has been disassembled and consolidated into a segmented cylindrical canister with rods from either 12 pressurized-water-reactor or 30 boiling-water-reactor assemblies per package for emplacement during Phase 2. 30 refs., 61 figs., 30 tabs.

  4. Iron Oxide Waste Form for Stabilizing 99Tc

    SciTech Connect

    Um, Wooyong; Chang, Hyun-Shik; Icenhower, Jonathan P.; Lukens, Wayne W.; Serne, R. Jeffrey; Qafoku, Nikolla; Kukkadapu, Ravi K.; Westsik, Joseph H.

    2012-06-09

    Crystals of goethite were synthesized with reduced technetium [{sup 99}Tc(IV)] incorporated within the solid lattice. The presence of {sup 99}Tc(IV) as a substituting cation in the matrix and 'armoring' by an additional layer of precipitated goethite isolated the reduced {sup 99}Tc(IV) from oxidizing agents. These products were used to make monolithic pellets to quantify an effective diffusion coefficient for {sup 99}Tc from goethite waste form contacted with a synthetic Hanford IDF (integrated disposal facility) pore water solution (pH = 7.2, I = 0.05 M) at room temperature for up to 120 days in static reactors. XANES analysis of the goethite solids recovered post-run demonstrated that the {sup 99}Tc in the goethite crystals remains in the reduced {sup 99}Tc(IV) state. The slow release of pertechnetate concentration with time in the static experiments with the monolith followed a square root of time dependence, consistent with diffusion control for {sup 99}Tc release. An apparent diffusion coefficient of 6.15 x 10{sup -11} cm{sup 2}/s was calculated for the {sup 99}Tc-goethite pellet sample and the corresponding leaching index (LI) was 10.2. The results of this study indicate that technetium can be immobilized in a stable, low-cost Fe oxide matrix that is easy to fabricate and these findings can be useful in designing long-term solutions for nuclear waste disposal.

  5. Iron oxide waste form for stabilizing 99Tc

    NASA Astrophysics Data System (ADS)

    Um, Wooyong; Chang, Hyunshik; Icenhower, Jonathan P.; Lukens, Wayne W.; Jeffrey Serne, R.; Qafoku, Nik; Kukkadapu, Ravi K.; Westsik, Joseph H.

    2012-10-01

    Crystals of goethite were synthesized with reduced technetium [99Tc(IV)] incorporated within the solid lattice. The presence of 99Tc(IV) as a substituting cation in the matrix and "armoring" by an additional layer of precipitated goethite isolated the reduced 99Tc(IV) from oxidizing agents. These products were used to make monolithic pellets to quantify an effective diffusion coefficient for 99Tc from goethite waste form contacted with a synthetic Hanford IDF (Integrated Disposal Facility) pore water solution (pH = 7.2 and I = 0.05 M) at room temperature for up to 120 days in static reactors. XANES analysis of the goethite solids recovered post-run demonstrated that the 99Tc in the goethite crystals remains in the reduced 99Tc(IV) state. The slow release of pertechnetate concentration with time in the static experiments with the monolith followed a square root of time dependence, consistent with diffusion control for 99Tc release. An apparent diffusion coefficient of 6.15 × 10-11 cm2/s was calculated for the 99Tc-goethite pellet sample and the corresponding leaching index (LI) was 10.2. The results of this study indicate that technetium can be immobilized in a stable, low-cost Fe oxide matrix that is easy to fabricate and these findings can be useful in designing long-term solutions for nuclear waste disposal.

  6. Iron oxide nanoparticles embedded in activated carbons prepared from hydrothermally treated waste biomass.

    PubMed

    Hao, Wenming; Björkman, Eva; Yun, Yifeng; Lilliestråle, Malte; Hedin, Niklas

    2014-03-01

    Particles of iron oxide (Fe3O4 ; 20–40 nm) were embedded within activated carbons during the activation of hydrothermally carbonized (HTC) biomasses in a flow of CO2. Four different HTC biomass samples (horse manure, grass cuttings, beer production waste, and biosludge) were used as precursors for the activated carbons. Nanoparticles of iron oxide formed from iron catalyst included in the HTC biomasses. After systematic optimization, the activated carbons had specific surface areas of about 800 m2g1. The pore size distributions of the activated carbons depended strongly on the degree of carbonization of the precursors. Activated carbons prepared from highly carbonized precursors had mainly micropores, whereas those prepared from less carbonized precursors contained mainly mesopores. Given the strong magnetism of the activated carbon–nano-Fe3O4 composites, they could be particularly useful for water purification.

  7. Bioleaching of zinc and iron from steel plant waste using Acidithiobacillus ferrooxidans.

    PubMed

    Bayat, Oktay; Sever, Efsun; Bayat, Belgin; Arslan, Volkan; Poole, Colin

    2009-01-01

    The bacterial leaching of zinc and iron from solid wastes at the Isdemir iron and steel plant has been investigated using Acidithiobacillus ferrooxidans as the bacterial agent. The effects of a range of operational parameters, including particle size, solids concentration and pH, on the efficiency of the bioleaching process were investigated. In each test, several variables were determined to assess the efficiency of leaching, including slurry pH and redox potential, temperature, bacteria population and concentrations of zinc and iron in solution. Experimental results demonstrated that pulp solids concentration, slurry pH and solids particle size were all important parameters in the bacterial leaching process. Maximum extraction was achieved at pH values around 1.3 and a solids concentration of 1% w/v, with 35% of the Zn content and 37% of the Fe being dissolved.

  8. Investigating the Utility of Iron Ore Waste in Preparing Non-fired Bricks

    NASA Astrophysics Data System (ADS)

    Lamani, Shreekant R.; Mangalpady, Aruna; Vardhan, Harsha

    2016-10-01

    Iron ore waste is a major problem for mine owners due to the difficulty involved in its storage, handling and other environmental related issues. An alternative solution to this is utilisation of iron ore waste (IOW) as some value added product in construction industry. An attempt has been made in this paper in examining the possibility of making non-fired bricks from iron ore waste with some additives like cement and fly-ash. Each of the additives were mixed with IOW in different ratios and different sets of bricks were prepared. The prepared IOW bricks were cured for 7, 14, 21 and 28 days and their respective compressive strength and percentage of water absorption were determined. The results show that IOW bricks prepared with 9% and above cement and with 28 days of curing are suitable for brick making and meet the IS specifications. It was also observed that the weight of the prepared bricks with 9% cement with 28 days of curing varies between 2.35 and 2.45 kg whereas the weight of compressed fire clay bricks varies from 2.80 to 2.89 kg. Results also show that the cost of bricks prepared with cement ranging from 9 to 20% is comparable to that of commercially available compressed bricks.

  9. Utilization of waste polyethylene terephthalate as a reducing agent in the reduction of iron ore composite pellets

    NASA Astrophysics Data System (ADS)

    Polat, Gökhan; Birol, Burak; Sarıdede, Muhlis Nezihi

    2014-08-01

    The increasing consumption of plastics inevitably results in increasing amounts of waste plastics. Because of their long degradation periods, these wastes negatively affect the natural environment. Numerous studies have been conducted to recycle and eliminate waste plastics. The potential for recycling waste plastics in the iron and steel industry has been underestimated; the high C and H contents of plastics may make them suitable as alternative reductants in the reduction process of iron ore. This study aims to substitute plastic wastes for coal in reduction melting process and to investigate their performance during reduction at high temperature. We used a common type of waste plastic, polyethylene terephthalate (PET), because of its high carbon and hydrogen contents. Composite pellets containing PET wastes, coke, and magnetite iron ore were reduced at selected temperatures of 1400 and 1450°C for reduction time from 2 to 10 min to investigate the reduction melting behavior of these pellets. The results showed that an increased temperature and reduction time increased the reduction ratio of the pellets. The optimum experimental conditions for obtaining metallic iron (iron nuggets) were reduction at 1450°C for 10 min using composite pellets containing 60% PET and 40% coke.

  10. 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.

  11. Researches on the production of self-reducing briquettes from waste containing iron and carbon

    NASA Astrophysics Data System (ADS)

    Ardelean, E.; Ardelean, M.; Hǎrǎu, C.; Crişan, E.

    2014-03-01

    The extension of the raw material basis for the steel making industry represents a priority nowadays, within the context of sustainable development. The issue is of utmost importance particularly now when the environment legislation sets strict dumping conditions on the one hand and, on the other hand, when huge quantities of waste, already deposited in dumps and ponds raise serious problems in terms of meeting such conditions. The paper introduces some researches on exploiting waste containing iron and carbon in steel making; this powdery and small-grain waste is to be processed into briquettes. The self-reducing briquettes can be used in steel elaboration in electric arc furnaces, replacing the scrap, which is scarce.

  12. Less Waste in the First Place. Six Lessons on Packaging and Source Reduction.

    ERIC Educational Resources Information Center

    Flexible Packaging Educational Foundation, Washington, DC.

    The six lessons contained in this guide provide students with a basic understanding of the scientific and environmental principles that govern modern package design and development. By completing the activities that accompany each of the lesson plans, students will be able to: (1) define a package and describe its functions; (2) differentiate…

  13. Intact and Degraded Criticality Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package

    SciTech Connect

    L.M. Montierth

    2000-09-15

    The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the U.S. Department of Energy's (DOE) Shippingport Light Water Breeder Reactor (SP LWBR) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP), which is to be placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (K{sub eff}) for intact- and degraded-mode internal configurations of the codisposal WP containing Shippingport LWBR seed-type assemblies. The results of this calculation will be used to evaluate criticality issues and support the analysis that is planed to be performed to demonstrate the viability of the codisposal concept for the MGR. This calculation is associated with the waste package design and was performed in accordance with the DOE SNF Analysis Plan for FY 2000 (See Ref. 22). The document has been prepared in accordance with the Administrative Procedure AP-3.12Q, Calculations (Ref. 23).

  14. Innovative Soft-Sided Waste Packaging System Implementation at a Small Department of Energy Environmental Restoration/Waste Management (ER/WM) Site

    SciTech Connect

    Wolf, J.

    2002-02-28

    Weiss Associates (WA) performs a broad range of environmental restoration/waste management (ER/WM) activities for the U.S. Department of Energy (DOE) at the former Laboratory for Energy-Related Health Research (LEHR), University of California, Davis (UC Davis). Over the last three years, the LEHR ER/WM program transitioned from a baseline packaging system of steel, 2.7 cubic meter (3.5-cubic yard) B-25 boxes to a 7.0 cubic meter (9.1-cubic yard) soft-sided container (Lift Liner) system. The transition increased efficiencies in processing, packaging, and storage, and when combined with decreased procurement costs, achieved a $402,000 cost savings (Table I). Additional disposal costs between $128,600 and $182,600 were avoided by minimizing void space. Future cost savings by the end of fiscal year 2003 are projected between $250,640 and $1,003,360.

  15. Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

    SciTech Connect

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

    This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs.

  16. Report to Congress on the potential use of lead in the waste packages for a geologic repository at Yucca Mountain, Nevada

    SciTech Connect

    1989-12-01

    In the Report of the Senate Committee on Appropriations accompanying the Energy and Water Appropriation Act for 1989, the Committee directed the Department of Energy (DOE) to evaluate the use of lead in the waste packages to be used in geologic repositories for spent nuclear fuel and high-level waste. The evaluation that was performed in response to this directive is presented in this report. This evaluation was based largely on a review of the technical literature on the behavior of lead, reports of work conducted in other countries, and work performed for the waste-management program being conducted by the DOE. The initial evaluation was limited to the potential use of lead in the packages to be used in the repository. Also, the focus of this report is post closure performance and not on retrievability and handling aspects of the waste package. 100 refs., 8 figs., 15 tabs.

  17. Effect of controlled potential on SCC of nuclear waste package container materials

    SciTech Connect

    Lum, B Y; Roy, A K; Spragge, M K

    1999-08-01

    The slow-strain-rate (SSR) test technique was used to evaluate the susceptibility of Titanium (Ti) Gr-7 (UNS R52400) and Ti Gr-12 (UNS R53400) to stress corrosion cracking (SCC). Ti Gr-7 and Ti Gr-12 are two candidate container materials for the multi-barrier package for nuclear waste. The tests were done in a deaerated 90 C acidic brine (pH {approx} 2.7) containing 5 weight percent (wt%) sodium chloride (NaCl) using a strain rate of 3.3 x 10{sup -6} sec{sup -1}. Before being tested in the acidic brine, specimens of each alloy were pulled inside the test chamber in the dry condition at ambient temperature. Then while in the test solution, specimens were strained under different cathodic (negative) controlled electrochemical potentials. These controlled potentials were selected based on the corrosion potential measured in the test solution before the specimens were strained. Results indicate that the times to failure (TTF) for Ti Gr-12 were much shorter than those for Ti Gr-7. Furthermore, as the applied potential became more cathodic, Ti Gr-12 showed reduced ductility in terms of percent reduction in area (%RA) and true fracture stress ({sigma}{sub f}). In addition, TTF and percent elongation (%El) reached the minimum values when Ti Gr-12 was tested under an impressed potential of -1162 mV. However, for Ti Gr-7, all these ductility parameters were not significantly influenced by the changes in applied potential. In general, the results of hydrogen analysis by secondary ion mass spectrometry (SIMS) showed increased hydrogen concentration at more cathodic controlled potentials. Optical microscopy and scanning electron microscopy (SEM) were used to evaluate the morphology of cracking both at the primary fracture face and the secondary cracks along the gage section of the broken tensile specimen. Transgranular secondary cracks were observed in both alloys possibly resulting from the formation of brittle titanium hydrides due to cathodic charging. The primary fracture

  18. Risk-informed criticality analysis as applied to waste packages subject to a subsurface igneous intrusion

    NASA Astrophysics Data System (ADS)

    Kimball, Darby Suzan

    branches of an event. This method of applying PRA techniques to criticality safety is demonstrated using the example of waste packages in an underground geologic repository during a volcanic event. It is concluded that the current design does not provide adequate subcritical assurance, and recommended that future design modifications focus on mitigating chemical degradation of fuel and metals.

  19. Removal of chromium from synthetic plating waste by zero-valent iron and sulfate-reducing bacteria.

    PubMed

    Guha, Saumyen; Bhargava, Puja

    2005-01-01

    Experiments were conducted to evaluate the potential of zero-valent iron and sulfate-reducing bacteria (SRB) for reduction and removal of chromium from synthetic electroplating waste. The zero-valent iron shows promising results as a reductant of hexavalent chromium (Cr+6) to trivalent chromium (Cr+3), capable of 100% reduction. The required iron concentration was a function of chromium concentration in the waste stream. Removal of Cr+3 by adsorption or precipitation on iron leads to complete removal of chromium from the waste and was a slower process than the reduction of Cr+6. Presence SRB in a completely mixed batch reactor inhibited the reduction of Cr+6. In a fixed-bed column reactor, SRB enhanced chromium removal and showed promising results for the treatment of wastes with low chromium concentrations. It is proposed that, for waste with high chromium concentration, zero-valent iron is an efficient reductant and can be used for reduction of Cr+6. For low chromium concentrations, a SRB augmented zero-valent iron and sand column is capable of removing chromium completely.

  20. Phytoremediation: a novel approach for utilization of iron-ore wastes.

    PubMed

    Mohanty, Monalisa; Dhal, Nabin Kumar; Patra, Parikshita; Das, Bisweswar; Reddy, Palli Sita Rama

    2010-01-01

    Large quantities of iron-ore tailings are being generated annually in the world from mining and processing of iron ores. It has been estimated that around 10-15% of the iron ore mined in India has remained unutilized and discarded as slimes during mining and subsequent processing. Soil contamination resulting from mining activities affects surrounding flora and fauna and presents a large clean-up challenge to the mining industry. Innovative new methodologies have been proposed and among the most promising are those that rely on new phytoremediation technology. In this paper we address and review the status of phytoremediation as a technology to reduce and control contaminated mine wastes. Several different approaches and different plant species are used to remove environmentally toxic metals from mine waste sites. Such approaches have the objective of restoring mining waste sites to human and animal use, or at least, to curtail or eliminate the off-site movement of toxic entities that potentially could reach humans. How well phytoremediation performs as an alternative soil restoration technology depends on several factors, including the composition of soil, toxicity level of the contaminant, degree to which plant species fit natural local growth patterns and type of concentration of metal/contaminant in such plants. Phytoremediation has opened prospects for less costly, yet practicable approaches to clean-up contaminated waste sites, particularly those associated with mineral extraction mining. We discuss several plant species that are capable of phytoextracting and/or phytostabilizing harmful elements from contaminated soil and water; such processes are prospectively effective for addressing waste problems that derive from mining and processing activities, as well as those that derive from mitigating the threat posed by waste that surrounds mining sites. Unfortunately, phytoremediation is still in the embryonic stage, and more research is needed to find the plant

  1. Evaluation and compilation of DOE waste package test data; Biannual report, February 1989--July 1989: Volume 7

    SciTech Connect

    Interrante, C.G.; Fraker, A.C.; Escalante, E.

    1991-12-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, February through July 1989. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Outlines for planned interpretative reports on the topics of aqueous corrosion of copper, mechanisms of stress corrosion cracking and internal failure modes of Zircaloy cladding are included. For the publications reviewed during this reporting period, short discussions are given to supplement the completed reviews and evaluations. Included in this report is an overall review of a 1984 report on glass leaching mechanisms, as well as reviews for each of the seven chapters of this report.

  2. Waste minimization assessment for a manufacturer of iron castings and fabricated sheet metal parts

    SciTech Connect

    Fleischman, M.; Harris, J.J.; Handmaker, A.; Looby, G.P.

    1995-08-01

    The U.S. Environmental Protection Agency (EPA) has funded a pilot project to assist small and medium-size manufacturers who want to minimize their generation of waste but who lack the expertise to do so. Waste Minimization Assessment Centers (WMACs) were established at selected universities and procedures were adapted from the EPA Waste Minimization Opportunity Assessment Manual. That document has been superseded by the Facility Pollution Prevention Guide. The WMAC team at the University of Louisville performed an assessment at a plant that manufactures iron castings and fabricated sheet metal parts. Foundry operations include mixing and mold formation, core making, metal pouring, shakeout, finishing, and painting. Cutting, shaping, and welding are the principal metal fabrication operations. The team`s report, detailing findings and recommendations indicated that paint-related wastes are generated in large quantities, and that significant waste reduction and cost savings could be realized by installing a dry powder coating system or by replacing conventional air spray paint guns with high-volume low-pressure spray guns. This research brief was developed by the principal investigators and EPA`s National Risk Management Research Laboratory, Cincinnati, OH, to announce key findings of an ongoing research project that is fully documented in a separate report of the same title available from University City Science Center.

  3. Potential for the localized corrosion of alloy 22 Waste Packages in Multiple-Salt Deliquescent Brines in the Yucca Mountain Repository

    SciTech Connect

    King, F.; Arthur, R.; Apted, M.; Kessler, J.H.

    2007-07-01

    It has been postulated that the deliquescence of multiple-salt systems in dust deposits and the consequent localized corrosion in high-temperature brines could lead to premature failure of the Alloy 22 waste packages in the Yucca Mountain repository. EPRI has developed a decision tree approach to determine if the various stages leading to waste package failure are possible and whether the safety of the repository system could be compromised as a result. Through a series of arguments, EPRI has shown that it is highly unlikely that the multiple-salt deliquescent brines will form in the first place and, even if they did, that they would not be thermodynamically stable, that the postulated brines are not corrosive and would not lead to the initiation of localized corrosion of Alloy 22, that even if localized corrosion did initiate that the propagation would stifle and cease long before penetration of the waste package outer barrier, and that even if premature waste package failures did occur from this cause that the safety of the overall system would not be compromised. EPRI concludes, therefore, that the postulated localized corrosion of the waste packages due to high-temperature deliquescent brines is neither a technical nor a safety issue of concern for the Yucca Mountain repository. (authors)

  4. Novel method of reducing transient emissions from rotary-kiln incinerators through modified waste packaging. Rept. for Mar 90-May 92

    SciTech Connect

    Lemieux, P.M.; Linak, W.P.; Wendt, J.O.L.

    1992-01-01

    The paper gives results of tests on a 73 kW pilot-scale rotary kiln incinerator simulator to examine the effect of modified waste packaging on the ability of the incineration system to respond to transients imposed due to batch charging of volatile liquid surrogate hazardous wastes bound on corncob sorbents. Normally, the waste container ruptures soon after its introduction into the kiln, rapidly releasing any contained volatile materials. This rapid release of combustibles has the potential to locally deplete available oxygen supplied by the main burner. This can lead to a transient puff of unburned material that must be dealt with by downstream equipment, such as an afterburner. Innovative waste packaging methods were used to delay the time required for waste vaporization, which significantly decreased the measured amount of unburned material exiting the kiln. This concept has advantageous implications for rotary kiln incinerators as a way to both reduce transient emissions and increase the feed rate of containerized liquids.

  5. Data Package for Secondary Waste Form Down-Selection—Cast Stone

    SciTech Connect

    Serne, R. Jeffrey; Westsik, Joseph H.

    2011-09-05

    Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations and leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.

  6. Stress Corrosion Cracking of the Drip Shield, The Waste Package Outer Barrier and the Stainless Steel Structural Material

    SciTech Connect

    C. Stephen

    2000-04-17

    One of the potential failure modes of the drip shield (DS), the waste package (WP) outer barrier, and the stainless structural material is the initiation and propagation of stress corrosion cracking (SCC) induced by the WP environment and various types of stresses that can develop in the DSs or the WPs. For the current design of the DS and WP, however, the DS will be excluded from the SCC evaluation because stresses that are relevant to SCC are insignificant in the DS. The major sources of stresses in the DS are loadings due to backfill and earthquakes. These stresses will not induce SCC because the stress caused by backfill is generally compressive stress and the stress caused by earthquakes is temporary in nature. The 316NG stainless steel inner barrier of the WP will also be excluded from the SCC evaluation because the SCC performance assessment will not take credit from the inner barrier. Therefore, the purpose of this document is to provide a detailed description of the process-level models that can be applied to assess the performance of the material (i.e., Alloy 22) used for the WP outer barrier subjected to the effects of SCC. As already mentioned in the development plan for the WP PMR (CRWMS M and O 1999e), this Analyses and Models Report (AMR) is to serve as a feed to the Waste Package Degradation (WPD) Total System Performance Assessment (TSPA) and Process Model Report (PMR).

  7. Closed circuit recovery of copper, lead and iron from electronic waste with citrate solutions.

    PubMed

    Torres, Robinson; Lapidus, Gretchen T

    2017-02-01

    An integral closed circuit hydrometallurgical process is presented for base metal recovery from electronic waste. The leaching medium consists of a sodium citrate solution, from which base metals are retrieved by direct electrowinning, and the barren solution is recycled back to the leaching stage. This leaching-electrowinning cycle was repeated four times. The redox properties of the fresh citrate solution, as well as the leach liquors, were characterized by cyclic voltammetry to determine adequate conditions for metal reduction, as well as to limit citrate degradation. The leaching efficiency of electronic waste, employing the same solution after four complete cycles was 71, 83 and 94% for copper, iron and lead, respectively, compared to the original leach with fresh citrate solution.

  8. Iron

    MedlinePlus

    Iron is a mineral that our bodies need for many functions. For example, iron is part of hemoglobin, a protein which carries ... It helps our muscles store and use oxygen. Iron is also part of many other proteins and ...

  9. Phytoavailability and fractions of iron and manganese in calcareous soil amended with composted urban wastes.

    PubMed

    Gallardo-Lara, Francisco; Azcón, Mariano; Polo, Alfredo

    2006-01-01

    Little is known about the effects of applying composted urban wastes on the phytoavailability and distribution of iron (Fe) and manganese (Mn) among chemical fractions in soil. In order to study this concern several experiments in pots containing calcareous soil were carried out. The received treatments by adding separately two rates (20 and 80 Mg ha-1) of municipal solid waste (MSW) compost and/or municipal solid waste and sewage sludge (MSW-SS) co-compost. The cropping sequence was a lettuce crop followed by a barley crop. It was observed that treatments amended with composted urban wastes tended to promote slight increases in lettuce yield compared to the control. The highest Fe levels in lettuce were found when higher rates of MSW-SS co-compost were applied; these values were significant compared to those obtained in the other treatments. In all cases, the application of organic materials increased the concentration and uptake of Mn in lettuce compared to the control; however, these increases were significant only when higher rates of MSW compost were applied. The organic amendments had beneficial delayed effects on barley yields, showing, in most cases, significant increases compared to the control. In this context, treatments with MSW compost were found to be more effective than the equivalent treatments amended with MSW-SS co-compost. Compared to the control, composted urban wastes increased Fe concentration in straw and rachis, and decreased Fe concentration in barley grain. Similarly, a decreased concentration of Mn in the dry matter of barley crop grown in soils treated with composted urban wastes was observed.

  10. Extended producer responsibility for packaging wastes and WEEE - a comparison of implementation and the role of local authorities across Europe.

    PubMed

    Cahill, Rachel; Grimes, Sue M; Wilson, David C

    2011-05-01

    A comparison of the implementation of extended producer responsibility (EPR) to packaging waste and waste electrical and electronic equipment (WEEE) is presented for a representative sample of eleven European Union countries based on five indicators: stakeholders and responsibilities; compliance mechanisms; role of local authorities; financing mechanisms and merits and limitations, with four countries selected for more detailed case study analysis. Similarities, trends and differences in national systems are highlighted with particular focus on the role of local authorities and their relationship with obligated producers and the effect on the operation and success of each system. The national systems vary considerably in design, in terms of influence of pre-existing policy and systems, methods of achieving producer compliance (multiple or single collective schemes), fee structures, targets, waste stream prioritization and local authority involvement. Differing approaches are evident across all member states with respect to the role played by local authorities, responsibility apportioned to them, and the evolution of working relationships between obligated producers and municipalities. On the whole, EPR for packaging and WEEE has been successfully implemented throughout Europe in terms of Directive targets. It is, however, clear that the EPR systems currently in application across Europe differ primarily due to contrasting opinion on the legitimacy of local authorities as stakeholders and, in some cases, a fear on the part of industry of associated costs. Where local authorities have been engaged in the design and implementation of national systems, existing infrastructure used and defined roles established for producers and local authorities, results have been significantly more positive than in the cases where local authorities have had limited engagement.

  11. The radiation characteristics of the transport packages with vitrified high-level waste

    SciTech Connect

    Bogatov, S. A.; Mitenkova, E. F. Novikov, N. V.

    2015-12-15

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  12. The radiation characteristics of the transport packages with vitrified high-level waste

    NASA Astrophysics Data System (ADS)

    Bogatov, S. A.; Mitenkova, E. F.; Novikov, N. V.

    2015-12-01

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  13. Removal of pertechnetate from simulated nuclear waste streams using supported zerovalent iron

    SciTech Connect

    Darab, John; Amonette, Alexandra; Burke, Deborah; Orr, Robert; Ponder, Sherman; Schrick, Bettina; Mallouk, Thomas; Lukens, Wayne; Caulder, Dana; Shuh, David

    2007-07-11

    The application of nanoparticles of predominantly zerovalent iron (nanoiron), either unsupported or supported, to the separation and reduction of pertechnetate anions (TcO4-) from complex waste mixtures was investigated as an alternative approach to current waste-processing schemes. Although applicable to pertechnetate-containing waste streams in general, the research discussed here was directed at two specific potential applications at the U.S. Department of Energy's Hanford Site: (1) the direct removal of pertechnetate from highly alkaline solutions, typical of those found in Hanford tank waste, and (2) the removal of dilute pertechnetate from near-neutral solutions, typical of the eluate streams from commercial organic ion-exchange resins that may be used to remediate Hanford tank wastes. It was envisioned that both applications would involve the subsequent encapsulation of the loaded sorbent material into a separate waste form. A high surface area (>200 M2/g) base-stable, nanocrystalline zirconia was used as a support for nanoiron for tests with highly alkaline solutions, while a silica gel support was used for tests with near-neutral solutions. It was shown that after 24 h of contact time, the high surface area zirconia supported nanoiron sorbent removed about 50percent (K-d = 370 L/kg) of the pertechnetate from a pH 14 tank waste simulant containing 0.51 mM TCO4- and large concentrations of Na+, OH-, NO3-, and CO32- for a phase ratio of 360 L simulant per kg of sorbent. It was also shown that after 18 h of contact time, the silica-supported nanoiron removed>95percent pertechnetate from a neutral pH eluate simulant containing 0.076 mM TcO4_ for a phase ratio of 290 L/kg. It was determined that in all cases, nanoiron reduced the Tc(VII) to Tc(IV), or possibly to Tc(V), through a redox reaction. Finally, it was demonstrated that a mixture of 20 mass percent of the solid reaction products obtained from contacting zirconia- supported nanoiron with an alkaline

  14. Removal of Pertechnetate From Simulated Nuclear Waste Streams Using Supported Zerovalent Iron

    SciTech Connect

    Darab, J.G.; Amonette, A.B.; Burke, D.S.D.; Orr, R.D.; Ponder, S.M.; Schrick, B.; Mallouk, T.E.; Lukens, W.W.; Caulder, D.L.; Shuh, D.K.

    2009-06-02

    The application of nanoparticles of predominantly zerovalent iron (nanoiron), either unsupported or supported, to the separation and reduction of pertechnetate anions (TcO{sub 4{sup -}}) from complex waste mixtures was investigated as an alternative approach to current waste-processing schemes. Although applicable to pertechnetate-containing waste streams in general, the research discussed here was directed at two specific potential applications at the U.S. Department of Energy's Hanford Site: (1) the direct removal of pertechnetate from highly alkaline solutions, typical of those found in Hanford tank waste, and (2) the removal of dilute pertechnetate from near-neutral solutions, typical of the eluate streams from commercial organic ion-exchange resins that may be used to remediate Hanford tank wastes. It was envisioned that both applications would involve the subsequent encapsulation of the loaded sorbent material into a separate waste form. A high surface area (>200 m{sup 2}/g) base-stable, nanocrystalline zirconia was used as a support for nanoiron for tests with highly alkaline solutions, while a silica gel support was used for tests with near-neutral solutions. It was shown that after 24 h of contact time, the high surface area zirconia supported nanoiron sorbent removed about 50% (K{sub d} = 370 L/kg) of the pertechnetate from a pH 14 tank waste simulant containing 0.51 mM TcO{sub 4{sup -}} and large concentrations of Na{sup +}, OH{sup -}, NO{sub 3{sup -}}, and CO{sub 3{sup 2-}} for a phase ratio of 360 L simulant per kg of sorbent. It was also shown that after 18 h of contact time, the silica-supported nanoiron removed >95% pertechnetate from a neutral pH eluate simulant containing 0.076 mM TcO{sub 4{sup -}} for a phase ratio of 290 L/kg. It was determined that in all cases, nanoiron reduced the Tc(VII) to Tc(IV), or possibly to Tc(V), through a redox reaction. Finally, it was demonstrated that a mixture of 20 mass % of the solid reaction products obtained

  15. Geology Data Package for the Single-Shell Tank Waste Management Areas at the Hanford Site

    SciTech Connect

    Reidel, Stephen P.; Chamness, Mickie A.

    2007-12-14

    This data package discusses the geology of the single-shell tank (SST) farms and the geologic history of the area. The purpose of this report is to provide the most recent geologic information available for the SST farms. This report builds upon previous reports on the tank farm geology and Integrated Disposal Facility geology with information available after those reports were published.

  16. Geology Data Package for the Single-Shell Tank Waste Management Areas at the Hanford Site

    SciTech Connect

    Reidel, Steve P.; Chamness, Mickie A.

    2007-01-01

    This data package discusses the geology of the single-shell tank (SST) farms and the geologic history of the area. The focus of this report is to provide the most recent geologic information available for the SST farms. This report builds upon previous reports on the tank farm geology and Integrated Disposal Facility geology with information available after those reports were published.

  17. Vendor Assessment for the Waste Package Closure System (Yucca Mountain Project)

    SciTech Connect

    Shelton-Davis, C.V.

    2003-09-26

    The Idaho National Engineering and Environmental Laboratory (INEEL) has been tasked with developing, designing, constructing, and operating a full-scale prototype of the work package closure system. As a precursor to developing the conceptual design, all commercially available equipment was assessed to identify any existing technology gaps. This report presents the results of that assessment for all major equipment.

  18. Vendor Assessment for the Waste Package Closure System (Yucca Mtn. Project)

    SciTech Connect

    Colleen Shelton-Davis

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) has been tasked with developing, designing, constructing, and operating a full-scale prototype of the work package closure system. As a precursor to developing the conceptual design, all commercially available equipment was assessed to identify any existing technology gaps. This report presents the results of that assessment for all major equipment.

  19. Tank vapor sampling and analysis data package for tank 241-C-106 waste retrieval sluicing system process test phase III

    SciTech Connect

    LOCKREM, L.L.

    1999-08-13

    This data package presents sampling data and analytical results from the March 28, 1999, vapor sampling of Hanford Site single-shell tank 241-C-106 during active sluicing. Samples were obtained from the 296-C-006 ventilation system stack and ambient air at several locations. Characterization Project Operations (CPO) was responsible for the collection of all SUMMATM canister samples. The Special Analytical Support (SAS) vapor team was responsible for the collection of all triple sorbent trap (TST), sorbent tube train (STT), polyurethane foam (PUF), and particulate filter samples collected at the 296-C-006 stack. The SAS vapor team used the non-electrical vapor sampling (NEVS) system to collect samples of the air, gases, and vapors from the 296-C-006 stack. The SAS vapor team collected and analyzed these samples for Lockheed Martin Hanford Corporation (LMHC) and Tank Waste Remediation System (TWRS) in accordance with the sampling and analytical requirements specified in the Waste Retrieval Sluicing System Vapor Sampling and Analysis Plan (SAP) for Evaluation of Organic Emissions, Process Test Phase III, HNF-4212, Rev. 0-A, (LMHC, 1999). All samples were stored in a secured Radioactive Materials Area (RMA) until the samples were radiologically released and received by SAS for analysis. The Waste Sampling and Characterization Facility (WSCF) performed the radiological analyses. The samples were received on April 5, 1999.

  20. CH Packaging Operations Manual

    SciTech Connect

    Washington TRU Solutions LLC

    2005-06-13

    This procedure provides instructions for assembling the CH Packaging Drum payload assembly, Standard Waste Box (SWB) assembly, Abnormal Operations and ICV and OCV Preshipment Leakage Rate Tests on the packaging seals, using a nondestructive Helium (He) Leak Test.

  1. Adsorption and removal of arsenic from water by iron ore mining waste.

    PubMed

    Nguyen, Tien Vinh; Nguyen, Thi Van Trang; Pham, Tuan Linh; Vigneswaran, Saravanamuth; Ngo, Huu Hao; Kandasamy, J; Nguyen, Hong Khanh; Nguyen, Duc Tho

    2009-01-01

    There is a global need to develop low-cost technologies to remove arsenic from water for individual household water supply. In this study, a purified and enriched waste material (treated magnetite waste, TMW) from the Trai Cau's iron ore mine in the Thai Nguyen Province in Vietnam was examined for its capacity to remove arsenic. The treatment system was packed with TMW that consisted of 75% of ferrous-ferric oxide (Fe(3)O(4)) and had a large surface area of 89.7 m(2)/g. The experiments were conducted at a filtration rate of 0.05 m/h to treat groundwater with an arsenic concentration of 380 microg/L and iron, manganese and phosphate concentrations of 2.07 mg/L, 0.093 mg/L and 1.6 mg/L respectively. The batch experimental results show that this new material was able to absorb up to 0.74 mg arsenic/g. The results also indicated that the treatment system removed more than 90% arsenic giving an effluent with an arsenic concentration of less than 30 microg/L while achieving a removal efficiency of about 80% for Mn(2 + ) and PO(4) (3-). This could be a promising and cost-effective new material for capturing arsenic as well as other metals from groundwater.

  2. Combustion of sponge iron plant wastes -- Char and fly ash in FBC boilers

    SciTech Connect

    Rajavel, M.; Muthukrishnan, M.; Banerjee, M.; Natarajan, R.

    1997-12-31

    Coal based sponge iron plants generate large quantity of waste materials in the form of kiln char and fly ash. This material has a very little fuel value owing to the fact that it contains a large percentage of ash and it is almost free from volatiles. Added to this, wide size range of the kiln rejects makes it practically impossible to burn in conventional firing. However, it is realized that the overall economy of the sponge iron plant is likely to be greatly enhanced if the char, the fly ash and a considerable quantity of coal fines, separated in the coal preparation plant, can be utilized. The only possible way to gainfully use these fuels is by fluidized bed combustion. Extensive tests were conducted with these fuels in bubbling fluidized bed combustion test facilities at BHEL, Tiruchirapalli, India. The presentation highlights the experience gained with the above waste fuels in BHEL test facilities and discusses the findings which were subsequently used to design large size FBC boilers.

  3. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite.

    PubMed

    Kaufhold, Stephan; Hassel, Achim Walter; Sanders, Daniel; Dohrmann, Reiner

    2015-03-21

    Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na-bentonites compared to the Ca-bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe-silicate. Up to now it is not clear why and how the patina formed. It, however, may be relevant as a corrosion inhibitor.

  4. Optimizing and developing a continuous separation system for the wet process separation of aluminum and polyethylene in aseptic composite packaging waste.

    PubMed

    Yan, Dahai; Peng, Zheng; Liu, Yuqiang; Li, Li; Huang, Qifei; Xie, Minghui; Wang, Qi

    2015-01-01

    The consumption of milk in China is increasing as living standards rapidly improve, and huge amounts of aseptic composite milk packaging waste are being generated. Aseptic composite packaging is composed of paper, polyethylene, and aluminum. It is difficult to separate the polyethylene and aluminum, so most of the waste is currently sent to landfill or incinerated with other municipal solid waste, meaning that enormous amounts of resources are wasted. A wet process technique for separating the aluminum and polyethylene from the composite materials after the paper had been removed from the original packaging waste was studied. The separation efficiency achieved using different separation reagents was compared, different separation mechanisms were explored, and the impacts of a range of parameters, such as the reagent concentration, temperature, and liquid-solid ratio, on the separation time and aluminum loss ratio were studied. Methanoic acid was found to be the optimal separation reagent, and the suitable conditions were a reagent concentration of 2-4 mol/L, a temperature of 60-80°C, and a liquid-solid ratio of 30 L/kg. These conditions allowed aluminum and polyethylene to be separated in less than 30 min, with an aluminum loss ratio of less than 3%. A mass balance was produced for the aluminum-polyethylene separation system, and control technique was developed to keep the ion concentrations in the reaction system stable. This allowed a continuous industrial-scale process for separating aluminum and polyethylene to be developed, and a demonstration facility with a capacity of 50t/d was built. The demonstration facility gave polyethylene and aluminum recovery rates of more than 98% and more than 72%, respectively. Separating 1t of aluminum-polyethylene composite packaging material gave a profit of 1769 Yuan, meaning that an effective method for recycling aseptic composite packaging waste was achieved.

  5. Viability Assessment of a Repository at Yucca Mountain. Volume 2: Preliminary Design Concept for the Repository and Waste Package

    SciTech Connect

    1998-12-01

    This volume describes the major design features of the Monitored Geologic Repository. This document is not intended to provide an exhaustive, detailed description of the repository design. Rather, this document summarizes the major systems and primary elements of the design that are radiologically significant, and references the specific technical documents and design analyses wherein the details can be found. Not all portions of the design are at the same level of completeness. Highest priority has been given to assigning resources to advance the design of the Monitored Geologic Repository features that are important to radiological safety and/or waste isolation and for which there is no NRC licensing precedent. Those features that are important to radiological safety and/or waste isolation, but for which there is an NRC precedent, receive second priority. Systems and features that have no impact on radiological safety or waste isolation receive the lowest priority. This prioritization process, referred to as binning, is discussed in more detail in Section 2.3. Not every subject discussed in this volume is given equal treatment with regard to the level of detail provided. For example, less detail is provided for the surface facility design than for the subsurface and waste package designs. This different level of detail is intentional. Greater detail is provided for those functions, structures, systems, and components that play key roles with regard to protecting radiological health and safety and that are not common to existing nuclear facilities already licensed by NRC. A number of radiological subjects are not addressed in the VA, (e.g., environmental qualification of equipment). Environmental qualification of equipment and other radiological safety considerations will be addressed in the LA. Non-radiological safety considerations such as silica dust control and other occupational safety considerations are considered equally important but are not addressed in

  6. Remaining Sites Verification Package for the 100-D-2 Lead Sheeting Waste Site, Waste Site Reclassification Form 2007-030

    SciTech Connect

    L. M. Dittmer

    2008-03-19

    The 100-D-2 Lead Sheeting waste site was located approximately 50 m southwest of the 185-D Building and approximately 16 m north of the east/west oriented road. The site consisted of a lead sheet covering a concrete pad. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  7. Six month progress report on the Waste Package Project at the University of Nevada, Las Vegas, July 1991--January 1992: Management, quality assurance and overview

    SciTech Connect

    Ladkany, S.G.

    1991-12-31

    The progress of the waste package project at the University of Nevada, Las Vegas was the subject of this report. It covered aspects of management and quality assurance, container design, application of ASME Pressure Vessel Codes, structural analysis of containers, design of rock tunnels for storage, and heat transfer phenomena. (MB)

  8. Six month progress report on the Waste Package Project at the University of Nevada, Las Vegas, July 1991--January 1992: Management, quality assurance and overview

    SciTech Connect

    Ladkany, S.G.

    1991-01-01

    The progress of the waste package project at the University of Nevada, Las Vegas was the subject of this report. It covered aspects of management and quality assurance, container design, application of ASME Pressure Vessel Codes, structural analysis of containers, design of rock tunnels for storage, and heat transfer phenomena. (MB)

  9. ENVIRONMENTAL RESEARCH BRIEF: WASTE REDUCTION ACTIVITIES AND OPTIONS FOR A LAMINATOR OF PAPER AND CARDBOARD PACKAGES.

    EPA Science Inventory

    The U.S. Environmental Protection Agency (EPA) funded a project with the New Jersey Department of Environmental Protection and Energy (NJDEPE) to assist in conducting waste minimization assessments at thirty small- to medium-sized businesses in the state of New Jersey. One of th...

  10. Remaining Sites Verification Package for the 120-F-1 Glass Dump Waste Site, Waste Site Reclassification Form 2008-028

    SciTech Connect

    J. M. Capron

    2008-06-27

    The 120-F-1 waste site consisted of two dumping areas located 660 m southeast of the 105-F Reactor containing laboratory equipment and bottles, demolition debris, light bulbs and tubes, small batteries, small drums, and pesticide contaminated soil. It is probable that 108-F was the source of the debris but the material may have come from other locations within the 100-F Area. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  11. RATES OF IRON OXIDATION AND ARSENIC SORPTION DURING GROUND WATER-SURFACE WATER MIXING AT A HAZARDOUS WASTE SITE

    EPA Science Inventory

    The fate of arsenic discharged from contaminated ground water to a pond at a hazardous waste site is controlled, in part, by the rate of ferrous iron oxidation-precipitation and arsenic sorption. Laboratory experiments were conducted using site-derived water to assess the impact...

  12. Substitution potentials of recycled HDPE and wood particles from post-consumer packaging waste in Wood-Plastic Composites.

    PubMed

    Sommerhuber, Philipp F; Welling, Johannes; Krause, Andreas

    2015-12-01

    The market share of Wood-Plastic Composites (WPC) is small but expected to grow sharply in Europe. This raises some concerns about suitable wood particles needed in the wood-based panels industry in Europe. Concerns are stimulated by the competition between the promotion of wooden products through the European Bioeconomy Strategy and wood as an energy carrier through the Renewable Energy Directive. Cascade use of resources and valorisation of waste are potential strategies to overcome resource scarcity. Under experimental design conditions, WPC made from post-consumer recycled wood and plastic (HDPE) were compared to WPC made from virgin resources. Wood content in the polymer matrix was raised in two steps from 0% to 30% and 60%. Mechanical and physical properties and colour differences were characterized. The feasibility of using cascaded resources for WPC is discussed. Results indicate the technical and economic feasibility of using recycled HDPE from packaging waste for WPC. Based on technical properties, 30% recycled wood content for WPC is feasible, but economic and political barriers of efficient cascading of biomass need to be overcome.

  13. Degradation mode survey candidate titanium-base alloys for Yucca Mountain project waste package materials. Revision 1

    SciTech Connect

    Gdowski, G.E.

    1997-12-01

    The Yucca Mountain Site Characterization Project (YMP) is evaluating materials from which to fabricate high-level nuclear waste containers (hereafter called waste packages) for the potential repository at Yucca Mountain, Nevada. Because of their very good corrosion resistance in aqueous environments titanium alloys are considered for container materials. Consideration of titanium alloys is understandable since about one-third (in 1978) of all titanium produced is used in applications where corrosion resistance is of primary importance. Consequently, there is a considerable amount of data which demonstrates that titanium alloys, in general, but particularly the commercial purity and dilute {alpha} grades, are highly corrosion resistant. This report will discuss the corrosion characteristics of Ti Gr 2, 7, 12, and 16. The more highly alloyed titanium alloys which were developed by adding a small Pd content to higher strength Ti alloys in order to give them better corrosion resistance will not be considered in this report. These alloys are all two phase ({alpha} and {beta}) alloys. The palladium addition while making these alloys more corrosion resistant does not give them the corrosion resistance of the single phase {alpha} and near-{alpha} (Ti Gr 12) alloys.

  14. Recycling of iron foundry sand and glass waste as raw material for production of whiteware.

    PubMed

    Bragança, Saulo R; Vicenzi, Juliane; Guerino, Kareline; Bergmann, Carlos P

    2006-02-01

    The purpose of this study was to evaluate the production feasibility of triaxial whiteware using sand from cast iron moulds as a raw material instead of silica, and recycled glass in place of feldspar. Formulations were prepared using sand, glass waste, and white-firing clay such that only 50% of the composition was virgin material (clay). The ceramic bodies were formed by pressing and fired at different temperatures (between 1100 and 1300 degrees C). Specimens were characterized in terms of green density prior to firing; and their flexural strength, linear shrinkage, and water absorption were measured after firing. The microstructure was determined by scanning electron microscopy. Possible environmental impacts of this recycling process were also evaluated, through solubility and leaching tests, according to Brazilian standards. Gaseous emissions during the firing process were also analysed. The results showed that it is possible to produce triaxial ceramics by using such alternative raw materials.

  15. Analysis of the optimization possibilities to recover the powdery wastes containing iron and carbon

    NASA Astrophysics Data System (ADS)

    Popescu, Darius-Alexandru; Vilceanu, Lucia; Socalici, Ana

    2016-06-01

    Most industrial activities result in one or more secondary products and wastes besides the primary product, with a variety of uses. The iron & steel industry is highly energy intensive, but it is also a major source of environmental pollution with gases and dusts, especially the extractive branch. The researches aimed the recovery of the dust from the sintering plants and blast furnaces through the briquetting technology. Its recovery is required either for preventing the pollution or for reducing the consumption of raw materials. The mechanical properties are important for the quality of briquettes. We presented in this paper a series of mathematical correlations among the mechanical properties and the components of the briquetting batch, obtained using Excel spreadsheet and MATLAB programs. After analysing the results, we choose the optimal variation limits for the briquetting batch components.

  16. Ferric iron-bearing sediments as a mineral trap for CO2 sequestration: Iron reduction using sulfur-bearing waste gas

    USGS Publications Warehouse

    Palandri, J.L.; Kharaka, Y.K.

    2005-01-01

    We present a novel method for geologic sequestration of anthropogenic CO2 in ferrous carbonate, using ferric iron present in widespread redbeds and other sediments. Iron can be reduced by SO2 that is commonly a component of flue gas produced by combustion of fossil fuel, or by adding SO2 or H2S derived from other industrial processes to the injected waste gas stream. Equilibrium and kinetically controlled geochemical simulations at 120 bar and 50 and 100 ??C with SO2 or H2S show that iron can be transformed almost entirely to siderite thereby trapping CO2, and simultaneously, that sulfur can be converted predominantly to dissolved sulfate. If there is an insufficient amount of sulfur-bearing gas relative to CO2 as for typical flue gas, then some of the iron is not reduced, and some of the CO2 is not sequestered. If there is an excess of sulfur-bearing gas, then complete iron reduction is ensured, and some of the iron precipitates as pyrite or other solid iron sulfide, depending on their relative precipitation kinetics. Gas mixtures with insufficient sulfur relative to CO2 can be used in sediments containing Ca, Mg, or other divalent metals capable of precipitating carbonate minerals. For quartz arenite with an initial porosity of 21% and containing 0.25 wt.% Fe2O3, approximately 0.7 g of CO2 is sequestered per kg of rock, and the porosity decrease is less than 0.03%. Sequestration of CO2 using ferric iron has the advantage of disposing of SO2 that may already be present in the combustion gas. ?? 2005 Published by Elsevier B.V.

  17. Enhanced high-solids anaerobic digestion of waste activated sludge by the addition of scrap iron.

    PubMed

    Zhang, Yaobin; Feng, Yinghong; Yu, Qilin; Xu, Zibin; Quan, Xie

    2014-05-01

    Anaerobic digestion of waste activated sludge usually requires pretreatment procedure to improve the bioavailability of sludge, which involves considerable energy and high expenditures. This study proposes a cost-effective method for enhanced anaerobic digestion of sludge without a pretreatment by directly adding iron into the digester. The results showed that addition of Fe(0) powder could enhance 14.46% methane yield, and Fe scrap (clean scrap) could further enhance methane yield (improving rate 21.28%) because the scrap has better mass transfer efficiency with sludge and liquid than Fe(0) powder. The scrap of Fe with rust (rusty scrap) could induce microbial Fe(III) reduction, which resulted in achieving the highest methane yield (improving rate 29.51%), and the reduction rate of volatile suspended solids (VSS) was also highest (48.27%) among Fe powder, clean scrap and rusty scrap. PCR-DGGE proved that the addition of rusty scrap could enhance diversity of acetobacteria and enrich iron-reducing bacteria to enhance degradation of complex substrates.

  18. Literature review of intrinsic actinide colloids related to spent fuel waste package release rates

    SciTech Connect

    Zhao, P.; Steward, S.A.

    1997-01-01

    Existence of actinide colloids provides an important mechanism in the migration of radionuclides and will be important in performance of a geologic repository for high-level nuclear waste. Actinide colloids have been formed during long-term unsaturated dissolution of spent fuel by groundwater. This article summarizes a literature search of actinide colloids. This report emphasizes the formation of intrinsic actinide colloids, because they would have the opportunity to form soon after groundwater contact with the spent fuel and before actinide-bearing groundwater reaches the surrounding geologic formations.

  19. Effect of ionizing radiation on the waste package environment; Revision 1

    SciTech Connect

    Reed, D.T.; Van Konynenburg, R.A.

    1991-05-01

    The radiolytic production of nitrogen oxides, nitrogen acids and ammonia are discussed in relation to the expected environment in a high-level waste repository that may be constructed at the Yucca Mountain site if it is found to be suitable. Both literature data and repository-relevant data are summarized for air-water vapor systems. The limiting cases of a dry air and a pure water vapor gas phase are also discussed. Design guidelines and recommendations, based solely on the potential consequence of radiation enhancement of corrosion, are given. 13 refs., 5 figs., 1 tab.

  20. Irradiation corrosion of waste package container materials in air/steam environment

    SciTech Connect

    Anantatmula, R.P.; Colburn, R.P.

    1991-11-01

    A set of corrosion tests of the candidate waste container materials for the Basalt Waste Isolation Project (BWIP) was conducted in a moist air environment simulating the repository preclosure condition. The tests were performed over a four-month period with air, saturated with moisture at 50{degree}C, passing through test chambers holding specimens at 150{degree}C and 250{degree}C. The test materials included A27 cast carbon steel, A387 ferritic steel, oxygen-free copper and cupronickel 90-10. Each material was tested both with and without basalt/bentonite packing material adjacent to the specimens. For each combination of material, temperature and packing condition, tests were conducted at 10,000 R/hr and 100 R/hr radiation levels. All four materials tested at 250{degree}C showed corrosion rates as much as an order of magnitude higher in the high radiation field when compared with similar tests without irradiation. Typically, the corrosion rates were lower at the lower dose rates, but were significantly higher at 100 R/hr compared to the case without radiation. At 150{degree}C only pure copper showed enhanced corrosion due to gamma radiation. The presence of packing material adjacent to the specimens had no consistent effect on corrosion rate at either temperature.

  1. Mineralogy and characterization of arsenic, iron, and lead in a mine waste-derived fertilizer

    SciTech Connect

    Williams, A.G.B.; Scheckel, K.G.; Tolaymat, T.; Impellitteri, C.A.

    2008-06-09

    The solid-state speciation of arsenic (As), iron (Fe), and lead (Pb) was studied in the mine waste-derived fertilizer Ironite using X-ray absorption spectroscopy, Moessbauer spectroscopy, and aging studies. Arsenic was primarily associated with ferrihydrite (60-70%), with the remainder found in arsenopyrite (30-40%). Lead was observed almost exclusively as anglesite (PbSO{sub 4}), with <1% observed as galena (PbS). The identification of As in oxidized Fe oxides and Pb as PbSO{sub 4} is in disagreement with the dominant reduced phases previously reported and suggests As and Pb contained within the mine waste-derived product are more bioavailable than previously considered. Aging studies in solution result in Ironite granules separating into two distinct fractions, an orange oxide precipitate and a crystalline fraction with a metallic luster. The orange oxide fraction contained As adsorbed/precipitated with ferrihydrite that is released into solution when allowed to equilibrate with water. The fraction with a metallic luster contained pyrite and arsenopyrite. A complete breakdown of arsenopyrite was observed in Ironite aged for 1 month in buffered deionized water. The observations from this study indicate As and Pb exist as oxidized phases that likely develop from the beneficiation and processing of mine tailings for commercial sale. The potential release of As and Pb has important implications for water quality standards and human health. Of particular concern is the quantity of As released from mine waste-derived products due to the new As regulation applied in 2006, limiting As levels to 10 {micro}g L{sup -1} in drinking water.

  2. The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

    SciTech Connect

    Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

    2003-09-15

    Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.

  3. Catalytic decomposition of tar derived from wood waste pyrolysis using Indonesian low grade iron ore as catalyst

    NASA Astrophysics Data System (ADS)

    Wicakso, Doni Rahmat; Sutijan, Rochmadi, Budiman, Arief

    2016-06-01

    Low grade iron ore can be used as an alternative catalyst for bio-tar decomposition. Compared to other catalysts, such as Ni, Rd, Ru, Pd and Pt, iron ore is cheaper. The objective of this research was to investigate the effect of using low grade iron ore as catalyst for tar catalytic decomposition in fixed bed reactor. Tar used in this experiment was pyrolysis product of wood waste while the catalyst was Indonesian low grade iron ore. The variables studied were temperatures between 500 - 600 °C and catalyst weight between 0 - 40 gram. The first step, tar was evaporated at 450 °C to produce tar vapor. Then, tar vapor was flowed to fixed bed reactor filled low grade iron ore. Gas and tar vapor from reactor was cooled, then the liquid and uncondensable gas were analyzed by GC/MS. The catalyst, after experiment, was weighed to calculate total carbon deposited into catalyst pores. The results showed that the tar components that were heavy and light hydrocarbon were decomposed and cracked within the iron ore pores to from gases, light hydrocarbon (bio-oil) and carbon, thus decreasing content tar in bio-oil and increasing the total gas product. In conclusion, the more low grade iron ore used as catalyst, the tar content in the liquid decrease, the H2 productivity increased and calorimetric value of bio-oil increased.

  4. Recovery of iron, carbon and zinc from steel plant waste oxides using the AISI-DOE postcombustion smelting technology

    SciTech Connect

    Sarma, B.; Downing, K.B.; Aukrust, E.

    1996-09-01

    This report describes a process to recover steel plant waste oxides to be used in the production of hot metal. The process flowsheet used at the pilot plant. Coal/coke breeze and iron ore pellets/waste oxides are charged into the smelting reactor. The waste oxides are either agglomerated into briquettes (1 inch) using a binder or micro-agglomerated into pellets (1/4 inch) without the use of a binder. The iron oxides dissolve in the slag and are reduced by carbon to produce molten iron. The gangue oxides present in the raw materials report to the slag. Coal charged to the smelter is both the fuel as well as the reductant. Carbon present in the waste oxides is also used as the fuel/reductant resulting in a decrease in the coal requirement. Oxygen is top blown through a central, water-cooled, dual circuit lance. Nitrogen is injected through tuyeres at the bottom of the reactor for stirring purposes. The hot metal and slag produced in the smelting reactor are tapped at regular intervals through a single taphole using a mudgun and drill system. The energy requirements of the process are provided by (i) the combustion of carbon to carbon monoxide, referred to as primary combustion and (ii) the combustion of CO and H{sub 2} to CO{sub 2} and H{sub 2}O, known as postcombustion.

  5. Environmental data package for ORNL Solid Waste Storage Area Four, the adjacent intermediate-level liquid waste transfer line, and the liquid waste pilot pit area

    SciTech Connect

    Davis, E.C.; Shoun, R.R.

    1986-09-01

    The Oak Ridge National Laboratory Remedial Action Program has determined through its review of past environmental studies that Solid Waste Storage Area Four (SWSA-4) continually releases radioactivity to White Oak Creek and therefore requires application of the site stabilization and remedial actions outlined under the 3004u provisions of the Resource Conservation and Recovery Act. Under these provisions, a Remedial Investigation/Feasibility Study (RI/FS) forms the basis for determining the extent of actions. This report assembles available historical and environmental data relative to the SWSA-4 waste area grouping (WAG), which includes the 9.3-ha SWSA-4 site, the adjacent abandoned intermediate-level liquid waste transfer line, and the experimental pilot pit area. The rationale for grouping these three waste management units into the SWSA-4 WAG is the fact that they each lie in the same hydrologic unit and share a common tributary to White Oak Creek. The results of this compilation demonstrate that although a considerable number of studies have been carried out in SWSA-4, needs such as installation of water quality wells and continued monitoring and reporting of hydrologic data still exist. These needs will become even more critical as the RI/FS process proceeds and remedial measures for the site are considered. Fewer studies have been carried out to characterize the extent of contamination at the waste transfer line and the pilot pit area. Alternatives for characterizing and stabilizing these two minor components of the SWSA-4 WAG are presented; however, extensive remedial actions do not appear to be warranted.

  6. Geologic Data Package for 2001 Immobilized Low-Activity Waste Performance Assessment

    SciTech Connect

    SP Reidel; DG Horton

    1999-12-21

    This database is a compilation of existing geologic data from both the existing and new immobilized low-activity waste disposal sites for use in the 2001 Performance Assessment. Data were compiled from both surface and subsurface geologic sources. Large-scale surface geologic maps, previously published, cover the entire 200-East Area and the disposal sites. Subsurface information consists of drilling and geophysical logs from nearby boreholes and stored sediment samples. Numerous published geological reports are available that describe the subsurface geology of the area. Site-specific subsurface data are summarized in tables and profiles in this document. Uncertainty in data is mainly restricted to borehole information. Variations in sampling and drilling techniques present some correlation uncertainties across the sites. A greater degree of uncertainty exists on the new site because of restricted borehole coverage. There is some uncertainty to the location and orientation of elastic dikes across the sites.

  7. Value added of mulberry paper waste by carboxymethylation for preparation a packaging film

    NASA Astrophysics Data System (ADS)

    Rachtanapun, P.; Kumthai, S.; Mulkarat, N.; Pintajam, N.; Suriyatem, R.

    2015-07-01

    Cellulose from mulberry paper waste was converted to carboxymethyl cellulose (CMCm) by etherification using chloroacetic acid and various sodium hydroxide (NaOH) concentrations (30-60 g/100 mL). The degree of substitution of the various CMCm materials produced ranged between 0.33-0.45. The chemical structure of cellulose and the synthesized CMCm materials was characterized using Fourier transform infrared spectroscopy. CMCm films were cast and tested. The tensile strength (TS) and percent elongation at break (EB) of the films were investigated. The highest TS (32.58 MPa) and EB (2.39%) were found using the 50g /100 mL NaOH-synthesized CMCm film. The effect of glycerol as a plasticizer on TS and EB of this CMCm film was also evaluated. Increasing the glycerol content decreased TS and increased EB of the CMCm films.

  8. Recharge Data Package for Hanford Single-Shell Tank Waste Management Areas

    SciTech Connect

    Fayer, Michael J.; Keller, Jason M.

    2007-09-24

    Pacific Northwest National Laboratory (PNNL) assists CH2M HILL Hanford Group, Inc., in its preparation of the Resource Conservation and Recovery Act (RCRA) Facility Investigation report. One of the PNNL tasks is to use existing information to estimate recharge rates for past and current conditions as well as future scenarios involving cleanup and closure of tank farms. The existing information includes recharge-relevant data collected during activities associated with a host of projects, including those of RCRA, the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), the CH2M HILL Tank Farm Vadose Zone Project, and the PNNL Remediation and Closure Science Project. As new information is published, the report contents can be updated. The objective of this data package was to use published data to provide recharge estimates for the scenarios being considered in the RCRA Facility Investigation. Recharge rates were estimated for areas that remain natural and undisturbed, areas where the vegetation has been disturbed, areas where both the vegetation and the soil have been disturbed, and areas that are engineered (e.g., surface barrier). The recharge estimates supplement the estimates provided by PNNL researchers in 2006 for the Hanford Site using additional field measurements and model analysis using weather data through 2006.

  9. Zero Valent Iron Significantly Enhances Methane Production from Waste Activated Sludge by Improving Biochemical Methane Potential Rather Than Hydrolysis Rate

    NASA Astrophysics Data System (ADS)

    Liu, Yiwen; Wang, Qilin; Zhang, Yaobin; Ni, Bing-Jie

    2015-02-01

    Anaerobic digestion has been widely applied for waste activated sludge (WAS) treatment. However, methane production from anaerobic digestion of WAS is usually limited by the slow hydrolysis rate and/or poor biochemical methane potential of WAS. This work systematically studied the effects of three different types of zero valent iron (i.e., iron powder, clean scrap and rusty scrap) on methane production from WAS in anaerobic digestion, by using both experimental and mathematical approaches. The results demonstrated that both the clean and the rusty iron scrap were more effective than the iron powder for improving methane production from WAS. Model-based analysis showed that ZVI addition significantly enhanced methane production from WAS through improving the biochemical methane potential of WAS rather than its hydrolysis rate. Economic analysis indicated that the ZVI-based technology for enhancing methane production from WAS is economically attractive, particularly considering that iron scrap can be freely acquired from industrial waste. Based on these results, the ZVI-based anaerobic digestion process of this work could be easily integrated with the conventional chemical phosphorus removal process in wastewater treatment plant to form a cost-effective and environment-friendly approach, enabling maximum resource recovery/reuse while achieving enhanced methane production in wastewater treatment system.

  10. Stimulating short-chain fatty acids production from waste activated sludge by nano zero-valent iron.

    PubMed

    Luo, Jingyang; Feng, Leiyu; Chen, Yinguang; Li, Xiang; Chen, Hong; Xiao, Naidong; Wang, Dongbo

    2014-10-10

    An efficient and green strategy, i.e. adding nano zero-valent iron into anaerobic fermentation systems to remarkably stimulate the accumulation of short-chain fatty acids from waste activated sludge via accelerating the solubilization and hydrolysis processes has been developed. In the presence of nano zero-valent iron, not only the short-chain fatty acids production was significantly improved, but also the fermentation time for maximal short-chain fatty acids was shortened compared with those in the absence of nano zero-valent iron. Mechanism investigations showed that the solubilization of sludge, hydrolysis of solubilized substances and acidification of hydrolyzed products were all enhanced by addition of nano zero-valent iron. Also, the general microbial activity of anaerobes and relative activities of key enzymes with hydrolysis and acidification of organic matters were improved than those in the control. 454 high-throughput pyrosequencing analysis suggested that the abundance of bacteria responsible for waste activated sludge hydrolysis and short-chain fatty acids production was greatly enhanced due to nano zero-valent iron addition.

  11. Zero Valent Iron Significantly Enhances Methane Production from Waste Activated Sludge by Improving Biochemical Methane Potential Rather Than Hydrolysis Rate

    PubMed Central

    Liu, Yiwen; Wang, Qilin; Zhang, Yaobin; Ni, Bing-Jie

    2015-01-01

    Anaerobic digestion has been widely applied for waste activated sludge (WAS) treatment. However, methane production from anaerobic digestion of WAS is usually limited by the slow hydrolysis rate and/or poor biochemical methane potential of WAS. This work systematically studied the effects of three different types of zero valent iron (i.e., iron powder, clean scrap and rusty scrap) on methane production from WAS in anaerobic digestion, by using both experimental and mathematical approaches. The results demonstrated that both the clean and the rusty iron scrap were more effective than the iron powder for improving methane production from WAS. Model-based analysis showed that ZVI addition significantly enhanced methane production from WAS through improving the biochemical methane potential of WAS rather than its hydrolysis rate. Economic analysis indicated that the ZVI-based technology for enhancing methane production from WAS is economically attractive, particularly considering that iron scrap can be freely acquired from industrial waste. Based on these results, the ZVI-based anaerobic digestion process of this work could be easily integrated with the conventional chemical phosphorus removal process in wastewater treatment plant to form a cost-effective and environment-friendly approach, enabling maximum resource recovery/reuse while achieving enhanced methane production in wastewater treatment system. PMID:25652244

  12. Zero valent iron significantly enhances methane production from waste activated sludge by improving biochemical methane potential rather than hydrolysis rate.

    PubMed

    Liu, Yiwen; Wang, Qilin; Zhang, Yaobin; Ni, Bing-Jie

    2015-02-05

    Anaerobic digestion has been widely applied for waste activated sludge (WAS) treatment. However, methane production from anaerobic digestion of WAS is usually limited by the slow hydrolysis rate and/or poor biochemical methane potential of WAS. This work systematically studied the effects of three different types of zero valent iron (i.e., iron powder, clean scrap and rusty scrap) on methane production from WAS in anaerobic digestion, by using both experimental and mathematical approaches. The results demonstrated that both the clean and the rusty iron scrap were more effective than the iron powder for improving methane production from WAS. Model-based analysis showed that ZVI addition significantly enhanced methane production from WAS through improving the biochemical methane potential of WAS rather than its hydrolysis rate. Economic analysis indicated that the ZVI-based technology for enhancing methane production from WAS is economically attractive, particularly considering that iron scrap can be freely acquired from industrial waste. Based on these results, the ZVI-based anaerobic digestion process of this work could be easily integrated with the conventional chemical phosphorus removal process in wastewater treatment plant to form a cost-effective and environment-friendly approach, enabling maximum resource recovery/reuse while achieving enhanced methane production in wastewater treatment system.

  13. EQ6 Calculation for Chemical Degradation of Shippingport LWBR (TH/U Oxide) Spent Nuclear Fuel Waste Packages

    SciTech Connect

    S. Arthur

    2000-09-14

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site. Because of the high content of fissile material in the SNF, the waste package (WP) design requires special consideration of the amount and placement of neutron absorbers and the possible loss of absorbers and SNF materials over geologic time. For some WPs, the outer shell corrosion-resistant material (CRM) and the corrosion-allowance inner shell may breach (Refs. 2 and 3), allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components and neutron absorbers from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing a Shippingport LWBR SNF seed assembly, and high-level waste (HLW) glass canisters arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which criticality control material, suggested for this WP design, will remain in the WP after corrosion/dissolution of the initial WP configuration (such that it can be effective in preventing criticality); (2) The extent to which fissile uranium and fertile thorium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this

  14. HOW THE ROCKY FLATS ENVIRONMENTAL TECHNOLOGY SITE DEVELOPED A NEW WASTE PACKAGE USING A POLYUREA COATING THAT IS SAFELY AND ECONOMICALLY ELIMINATING SIZE REDUCTION OF LARGE ITEMS

    SciTech Connect

    Dorr, Kent A.; Hogue, Richard S.; Kimokeo, Margaret K.

    2003-02-27

    One of the major challenges involved in closing the Rocky Flats Environmental Technology Site (RFETS) is the disposal of extremely large pieces of contaminated production equipment and building debris. Past practice has been to size reduce the equipment into pieces small enough to fit into approved, standard waste containers. Size reducing this equipment is extremely expensive, and exposes workers to high-risk tasks, including significant industrial, chemical, and radiological hazards. RFETS has developed a waste package using a Polyurea coating for shipping large contaminated objects. The cost and schedule savings have been significant.

  15. Technical assessment of processing plants as exemplified by the sorting of beverage cartons from lightweight packaging wastes.

    PubMed

    Feil, A; Thoden van Velzen, E U; Jansen, M; Vitz, P; Go, N; Pretz, T

    2016-02-01

    The recovery of beverage cartons (BC) in three lightweight packaging waste processing plants (LP) was analyzed with different input materials and input masses in the area of 21-50Mg. The data was generated by gravimetric determination of the sorting products, sampling and sorting analysis. Since the particle size of beverage cartons is larger than 120mm, a modified sampling plan was implemented and targeted multiple sampling (3-11 individual samplings) and a total sample size of respectively 1200l (ca. 60kg) for the BC-products and of about 2400l (ca. 120kg) for material-heterogeneous mixed plastics (MP) and sorting residue products. The results infer that the quantification of the beverage carton yield in the process, i.e., by including all product-containing material streams, can be specified only with considerable fluctuation ranges. Consequently, the total assessment, regarding all product streams, is rather qualitative than quantitative. Irregular operation conditions as well as unfavorable sampling conditions and capacity overloads are likely causes for high confidence intervals. From the results of the current study, recommendations can basically be derived for a better sampling in LP-processing plants. Despite of the suboptimal statistical results, the results indicate very clear that the plants show definite optimisation potentials with regard to the yield of beverage cartons as well as the required product purity. Due to the test character of the sorting trials the plant parameterization was not ideal for this sorting task and consequently the results should be interpreted with care.

  16. Waste Receiving and Packaging, Module 2A, Supplemental Design Requirements Document

    SciTech Connect

    Lamberd, D.L.; Boothe, G.F.; Hinkle, A.L.; Horgos, R.M.; LeClair, M.D.; Nash, C.R.; Ocampo, V.P.; Pauly, T.R.; Stroup, J.L.; Weingardt, K.M.

    1994-04-26

    The Supplemental Design Requirements Document (SDRD) is used to communicate plant design information from Westinghouse Hanford Company (WHC) to the US Department of Energy (DOE) and the cognizant Architect Engineer (A/E). Information in the SDRD serves two purposes: to convey design requirements that are too detailed for inclusion in a Functional Design Criteria (FDC) report; and to serve as a means of change control for design commitments in the Conceptual Design Report. The mission of WRAP 2A on the Hanford site is the treatment of contact handled low level mixed waste (MW) for final disposal. The overall systems engineering steps used to reach construction and operation of WRAP 2A are depicted in Figure 1. The WRAP 2A SDRD focuses on the requirements to address the functional analysis provided in Figure 1. This information is provided in sections 2 through 5 of this SDRD. The mission analysis and functional analysis are to be provided in a separate supporting document. The organization of sections 2 through 5 corresponds to the requirements identified in the WRAP 2A functional analysis.

  17. EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

    SciTech Connect

    P. Bernot

    2001-02-27

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to

  18. The feasibility of low-G grey solidification of nodular iron in the F-104 experimental furnace package

    NASA Technical Reports Server (NTRS)

    Curreri, P. A.; Smith, G. A.; Workman, G.

    1983-01-01

    The rationale for low-g experiments with cast iron and the need for solidification in the grey form during these experiments are reviewed. The factors which determine whether an iron melt will solidify grey or white are discussed. Cooling rate versus microstructure was studied for a nodular iron candidate material for F-104 low-g solidification. The study determined that low-g grey solidification, using the present F-104 furnace system, of the nodular iron composition studied is not feasible. Specimen microstructure strongly suggested that the F-104 furnace's gas cooling system was causing excessive localized chill resulting in the nucleation of the unwanted iron carbide phase. A change is suggested, in the quench system design, that could possibly overcome this problem.

  19. Calculation Package for the Analysis of Performance of Cells 1-6, with Underdrain, of the Environmental Management Waste Management Facility Oak Ridge, Tennessee

    SciTech Connect

    Gonzales D.

    2010-03-30

    This calculation package presents the results of an assessment of the performance of the 6 cell design of the Environmental Management Waste Management Facility (EMWMF). The calculations show that the new cell 6 design at the EMWMF meets the current WAC requirement. QA/QC steps were taken to verify the input/output data for the risk model and data transfer from modeling output files to tables and calculation.

  20. Borehole Data Package for 1998 Wells Installed at Single-Shell Tank Waste Management Area U

    SciTech Connect

    DG Horton; FN Hodges

    1999-03-23

    Two new Resource Conservation and Recovery Act (RCR4) groundwater monitoring wells were installed at the single-shell tank f- Waste Management Area (WMA) U in October 1998 in fi,dfillrnent of Tri-PaQy Agreement (Ecology 1996) milestone M-24-39. The wells are 299-W19-41 and 299-W19-42. Well 299-W19-41 is located east of the southeastern comer of the WMA and replaces downgradient well 299-W19-32. Well 299-W19-42 is located east of the WNIA near the northeastern comer and is a new downgradient monitoring well. The locations of all wells in the monitoring network are shown on Figure 1. The groundwater monitoring plan for WMA U (Caggiano and Goodwin 1991) describes the hydrogeology of the 200 West Area and WMA U. An Interim Change Notice to the groundwater monitoring plan provides justification for the new wells. The new wells were constructed to the specifications and requirements described in Washington Administrative Code (WAC) 173-160 and WAC 173-303. This document compiles Mormation on the drilling and Construction well development pump installation, and sedment testing applicable to wells 299-W19-41 and 299-W19-42. Appendix A contains the geologist's log, the Well Construction Summary Reportj and Well Summary Sheet (as-built diagram); Appendix B contains results of laboratory measurements of particle size distribution pm conductivity, and calcium carbonate and moisture contents; and Append~ C contains geophysical logs. Aquifer tests (slug tests) were pefiormed on both new wells. Results from the aquifer tests will be reported elsewhere. Additional documentation concerning well construction is on file with Bechtel Hanford, Inc., Richland Washington.

  1. Ultrasonic recovery of copper and iron through the simultaneous utilization of Printed Circuit Boards (PCB) spent acid etching solution and PCB waste sludge.

    PubMed

    Huang, Zhiyuan; Xie, Fengchun; Ma, Yang

    2011-01-15

    A method was developed to recover the copper and iron from Printed Circuit Boards (PCB) manufacturing generated spent acid etching solution and waste sludge with ultrasonic energy at laboratory scale. It demonstrated that copper-containing PCB spent etching solution could be utilized as a leaching solution to leach copper from copper contained PCB waste sludge. It also indicated that lime could be used as an alkaline precipitating agent in this method to precipitate iron from the mixture of acidic PCB spent etching solution and waste sludge. This method provided an effective technique for the recovery of copper and iron through simultaneous use of PCB spent acid solution and waste sludge. The leaching rates of copper and iron enhanced with ultrasound energy were reached at 93.76% and 2.07% respectively and effectively separated copper from iron. Followed by applying lime to precipitate copper from the mixture of leachate and rinsing water produced by the copper and iron separation, about 99.99% and 1.29% of soluble copper and calcium were settled as the solids respectively. Furthermore the settled copper could be made as commercial rate copper. The process performance parameters studied were pH, ultrasonic power, and temperature. This method provided a simple and reliable technique to recover copper and iron from waste streams generated by PCB manufacturing, and would significantly reduce the cost of chemicals used in the recovery.

  2. Scanning electron microscopy and magnetic characterization of iron oxides in solid waste landfill leachate.

    PubMed

    Huliselan, Estevanus Kristian; Bijaksana, Satria; Srigutomo, Wahyu; Kardena, Edwan

    2010-07-15

    Leachate sludge samples were taken from two municipal solid waste sites of Jelekong and Sarimukti in Bandung, Indonesia. Their magnetic mineralogy and granulometry were analyzed to discriminate the sources of magnetic minerals using X-ray diffraction (XRD), scanning electron microscopy equipped with energy-dispersive X-ray spectroscopy (SEM-EDX) and rock magnetism. SEM-EDX analyses infer that the main magnetic minerals in the leachate sludge are iron oxides. In terms of their morphology, the grains from Jelekong are mostly octahedral and angular, which are similar to the general shapes of magnetic grains from the local soils. The grains from Sarimukti, on the other hand, are dominated by imperfect spherule shapes suggesting the product of combustion processes. Hysteresis parameters verify that the predominant magnetic mineral in leachate sludge is low coercivity ferrimagnetic mineral such as magnetite (Fe(3)O(4)). Furthermore, comparisons of rock magnetic parameters show that the magnetic minerals of soil samples from Jelekong have higher degree of magnetic pedogenesis indicating higher proportion of superparamagnetic/ultrafine particles than those of soil samples from Sarimukti. The plot of susceptibilities ratio versus coercive force has a great potential to be used as a discriminating tool for determining the source of magnetic minerals.

  3. Enhanced anaerobic digestion of waste activated sludge digestion by the addition of zero valent iron.

    PubMed

    Feng, Yinghong; Zhang, Yaobin; Quan, Xie; Chen, Suo

    2014-04-01

    Anaerobic digestion is promising technology to recover energy from waste activated sludge. However, the sludge digestion is limited by its low efficiency of hydrolysis-acidification. Zero valent iron (ZVI) as a reducing material is expected to enhance anaerobic process including the hydrolysis-acidification process. Considering that, ZVI was added into an anaerobic sludge digestion system to accelerate the sludge digestion in this study. The results indicated that ZVI effectively enhanced the decomposition of protein and cellulose, the two main components of the sludge. Compared to the control test without ZVI, the degradation of protein increased 21.9% and the volatile fatty acids production increased 37.3% with adding ZVI. More acetate and less propionate are found during the hydrolysis-acidification with ZVI. The activities of several key enzymes in the hydrolysis and acidification increased 0.6-1 time. ZVI made the methane production raise 43.5% and sludge reduction ratio increase 12.2 percent points. Fluorescence in situ hybridization analysis showed that the abundances of hydrogen-consuming microorganisms including homoacetogens and hydrogenotrophic methanogens with ZVI were higher than the control, which reduced the H2 accumulation to create a beneficial condition for the sludge digestion in thermodynamics.

  4. Arsenic biotransformation in solid waste residue: comparison of contributions from bacteria with arsenate and iron reducing pathways.

    PubMed

    Tian, Haixia; Shi, Qiantao; Jing, Chuanyong

    2015-02-17

    Arsenic- and iron-reducing bacteria play an important role in regulating As redox transformation and mobility. The motivation of this study was to compare the contributions of different As- and Fe-reducing bacteria to As biotransformation. In this work, three bacteria strains with different functional genes were employed including Pantoea sp. IMH with the arsC gene, Alkaliphilus oremlandii OhILAs possessing the arrA gene, and Shewanella oneidensis MR-1, an iron reducer. The incubation results showed that Pantoea sp. IMH aerobically reduced 100% of As(V) released from waste residues, though total As release was not enhanced. Similarly, strain OhILAs anaerobically reduced dissolved As(V) but could not enhance As release. In contrast, strain MR-1 substantially enhanced As mobilization because of iron reduction, but without changing the As speciation. The formation of the secondary iron mineral pyrite in the MR-1 incubation experiments, as evidenced by the X-ray absorption near-edge spectroscopy (XANES) analysis, contributed little to the uptake of the freed As. Our results suggest that the arsC gene carriers mainly control the As speciation in the aqueous phase in aerobic environments, whereas in anaerobic conditions, the As speciation should be regulated by arrA gene carriers, and As mobility is greatly enhanced by iron reduction.

  5. Edible packaging materials.

    PubMed

    Janjarasskul, Theeranun; Krochta, John M

    2010-01-01

    Research groups and the food and pharmaceutical industries recognize edible packaging as a useful alternative or addition to conventional packaging to reduce waste and to create novel applications for improving product stability, quality, safety, variety, and convenience for consumers. Recent studies have explored the ability of biopolymer-based food packaging materials to carry and control-release active compounds. As diverse edible packaging materials derived from various by-products or waste from food industry are being developed, the dry thermoplastic process is advancing rapidly as a feasible commercial edible packaging manufacturing process. The employment of nanocomposite concepts to edible packaging materials promises to improve barrier and mechanical properties and facilitate effective incorporation of bioactive ingredients and other designed functions. In addition to the need for a more fundamental understanding to enable design to desired specifications, edible packaging has to overcome challenges such as regulatory requirements, consumer acceptance, and scaling-up research concepts to commercial applications.

  6. Packaging legislation. Objectives and consequences.

    PubMed

    Christmann, H

    1995-05-01

    The recently published Directive on packaging and packaging waste makes new demands on the industry. This article highlights the key areas and raises some of the issues that must be confronted in the future.

  7. IN-PACKAGE CHEMISTRY ABSTRACTION

    SciTech Connect

    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 EQ3/6 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.

  8. Solid waste reclamation and recycling: Packaging and containers. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1996-08-01

    The bibliography contains citations concerning techniques and management of packaging and container recycling. References discuss recycling of tin and aluminum cans, reverse vending machines, reusable packaging and containers, and the future of containers. Environmental aspects, government programs, and development of recycling markets are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  9. Solid waste reclamation and recycling: Packaging and containers. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1995-09-01

    The bibliography contains citations concerning techniques and management of packaging and container recycling. References discuss recycling of tin and aluminum cans, reverse vending machines, reusable packaging and containers, and the future of containers. Environmental aspects, government programs, and development of recycling markets are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  10. Enhancing methane production from waste activated sludge using a novel indigenous iron activated peroxidation pre-treatment process.

    PubMed

    Zhou, Xu; Wang, Qilin; Jiang, Guangming

    2015-04-01

    Methane production from anaerobic digestion of waste activated sludge (WAS) is limited by the slow hydrolysis rate and/or poor methane potential of WAS. This study presents a novel pre-treatment strategy based on indigenous iron (in WAS) activated peroxidation to enhance methane production from WAS. Pre-treatment of WAS for 30 min at 50mg H2O2/g total solids (dry weight) and pH 2.0 (iron concentration in WAS was 7 mg/g TS) substantially enhanced WAS solubilization. Biochemical methane potential tests demonstrated that methane production was improved by 10% at a digestion time of 16d after incorporating the indigenous iron activated peroxidation pre-treatment. Model-based analysis indicated that indigenous iron activated peroxidation pre-treatment improved the methane potential by 13%, whereas the hydrolysis rate was not significantly affected. The economic analysis showed that the proposed pre-treatment method can save the cost by $112,000 per year in a treatment plant with a population equivalent of 300,000.

  11. Packaging for Posterity.

    ERIC Educational Resources Information Center

    Sias, Jim

    1990-01-01

    A project in which students designed environmentally responsible food packaging is described. The problem definition; research on topics such as waste paper, plastic, metal, glass, incineration, recycling, and consumer preferences; and the presentation design are provided. (KR)

  12. Modeling of radionuclide and heavy metal sorption around low and high pH waste disposal sites at Oak Ridge, Tennessee: Classification review package

    SciTech Connect

    Saunders, J.A.; Toran, L.E.

    1994-10-01

    Modeling of mineral precipitation and metal sorption reactions using MINTEQA2 and the iron oxyhydroxide diffuse-layer model has provided insights into geochemical processes governing contaminant migration from low-level radioactive waste disposal sites at the US Department of Energy`s Oak Ridge National Laboratory and Y-12 Plant at Oak Ridge, Tennessee. Both acidic and basic nuclear-fuel reprocessing wastes, locally mixed with decontamination solvents, were disposed of in unlined trenches and lagoons. Model results show that as wastes move toward neutral pH due to reactions with surrounding soils and saprolite, mineral precipitation and sorption can limit the solubility of heavy metals and radionuclides. However, observed contaminant levels in monitoring wells indicate that at least locally, wastes are moving in faults and fractures and are not retarded by sorption reactions along such flow paths. Model results also support previous studies that have indicated organic complexing agents used in decontamination procedures can enhance radionuclide and heavy metal solubility when mixed with nuclear fuel reprocessing wastes. However, complex interactions between metal-organic complexes and mineral surfaces and natural organic matter, biodegradation, and fracture flow complicate the interpretation of contaminant mobility.

  13. Background studies in support of a feasibility assessment on the use of copper-base materials for nuclear waste packages in a repository in tuff

    SciTech Connect

    Van Konynenburg, R.A.; Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S.

    1990-06-01

    This report combines six work units performed in FY`85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs.

  14. [The Waste Package Project. Final report, July 1, 1995--February 27, 1996]: Volume 3, Stress study in faulted tunnel models by combined photoelastic measurements and finite element analysis

    SciTech Connect

    Huang, Yuping

    1995-12-01

    The aim of this part of the Nuclear Waste Package Project research at UNLV is to investigate the stresses in a model of a faulted mountain and the effect of the fault on the stability of drifts in a proposed High Level Nuclear Waste Repository. An investigation was performed to develop a proper technique for analyzing the stresses in and around three adjacent scaled tunnel models, along with the stress concentration factors resulting from the existence of a fault that penetrates two of the three tunnels, at an inclined angel of 44{degrees} to the horizontal plane. The results and experience gained from this investigation will be used in a future project in which a full-size repository drift and a penetrating fault will be modeled and analyzed.

  15. Proceedings of the 6th Annual Meeting for Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and WasteTreatment, Storage and Disposal Activities

    SciTech Connect

    Jardine, L J

    2005-06-30

    The sixth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held November 15-17, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, and Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 55 Russian attendees from 16 different Russian organizations and four non-Russian attendees from the US. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C. The 16 different Russian design, industrial sites, and scientific organizations in attendance included staff from Rosatom/Minatom, Federal Nuclear and Radiation Safety Authority of Russia (GOSATOMNADZOR, NIERA/GAN), All Russian Designing & Scientific Research Institute of Complex Power Technology (VNIPIET), Khlopin Radium Institute (KRI), A. A. Bochvar All Russian Scientific Research Institute of Inorganic Materials (VNIINM), All Russian & Design Institute of Production Engineering (VNIPIPT), Ministry of Atomic Energy of Russian Federation Specialized State Designing Institute (GSPI), State Scientific Center Research Institute of Atomic Reactors (RIAR), Siberian Chemical Combine Tomsk (SCC), Mayak PO, Mining Chemical Combine (MCC K-26), Institute of Biophysics (IBPh), Sverdlosk Scientific Research Institute of Chemical Machine Building (SNIIChM), Kurchatov Institute (KI), Institute of Physical Chemistry Russian Academy of Science (IPCh RAS) and Radon PO-Moscow. The four non-Russian attendees included

  16. Treatment of air pollution control residues with iron rich waste sulfuric acid: does it work for antimony (Sb)?

    PubMed

    Okkenhaug, Gudny; Breedveld, Gijs D; Kirkeng, Terje; Lægreid, Marit; Mæhlum, Trond; Mulder, Jan

    2013-03-15

    Antimony (Sb) in air pollution control (APC) residues from municipal solid waste incineration has gained increased focus due to strict Sb leaching limits set by the EU landfill directive. Here we study the chemical speciation and solubility of Sb at the APC treatment facility NOAH Langøya (Norway), where iron (Fe)-rich sulfuric acid (∼3.6M, 2.3% Fe(II)), a waste product from the industrial extraction of ilmenite, is used for neutralization. Antimony in water extracts of untreated APC residues occurred exclusively as pentavalent antimonate, even at low pH and Eh values. The Sb solubility increased substantially at pH<10, possibly due to the dissolution of ettringite (at alkaline pH) or calcium (Ca)-antimonate. Treated APC residues, stored anoxically in the laboratory, simulating the conditions at the NOAH Langøya landfill, gave rise to decreasing concentrations of Sb in porewater, occurring exclusively as Sb(V). Concentrations of Sb decreased from 87-918μgL(-1) (day 3) to 18-69μgL(-1) (day 600). We hypothesize that an initial sorption of Sb to Fe(II)-Fe(III) hydroxides (green rust) and eventually precipitation of Ca- and Fe-antimonates (tripuhyite; FeSbO4) occurred. We conclude that Fe-rich, sulfuric acid waste is efficient to immobilize Sb in APC residues from waste incineration.

  17. 40 CFR 262.30 - Packaging.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... APPLICABLE TO GENERATORS OF HAZARDOUS WASTE Pre-Transport Requirements § 262.30 Packaging. Before transporting hazardous waste or offering hazardous waste for transportation off-site, a generator must...

  18. 40 CFR 262.30 - Packaging.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... APPLICABLE TO GENERATORS OF HAZARDOUS WASTE Pre-Transport Requirements § 262.30 Packaging. Before transporting hazardous waste or offering hazardous waste for transportation off-site, a generator must...

  19. 40 CFR 262.30 - Packaging.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... APPLICABLE TO GENERATORS OF HAZARDOUS WASTE Pre-Transport Requirements § 262.30 Packaging. Before transporting hazardous waste or offering hazardous waste for transportation off-site, a generator must...

  20. 40 CFR 262.30 - Packaging.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... APPLICABLE TO GENERATORS OF HAZARDOUS WASTE Pre-Transport Requirements § 262.30 Packaging. Before transporting hazardous waste or offering hazardous waste for transportation off-site, a generator must...

  1. 40 CFR 262.30 - Packaging.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... APPLICABLE TO GENERATORS OF HAZARDOUS WASTE Pre-Transport Requirements § 262.30 Packaging. Before transporting hazardous waste or offering hazardous waste for transportation off-site, a generator must...

  2. Geochemical and mineralogical characterization of sulfur and iron in coal waste rock, Elk Valley, British Columbia, Canada.

    PubMed

    Essilfie-Dughan, Joseph; Hendry, M Jim; Dynes, James J; Hu, Yongfeng; Biswas, Ashis; Lee Barbour, S; Day, S

    2017-02-12

    Exposure of coal waste rock to atmospheric oxygen can result in the oxidation of sulfide minerals and the release of sulfate (SO4(2-)) and associated trace elements (e.g., Se, As, Cd, and Zn) to groundwaters and surface waters. Similarly, reduced iron minerals such as siderite, ankerite, and the sulfide, pyrite, present in the waste rock can also undergo oxidation, resulting in the formation of iron oxyhydroxides that can adsorb trace elements released from the oxidation of the sulfide minerals. Characterization and quantification of the distribution of sulfide and iron minerals, their oxidation products, as well as leaching rates are critical to assessing present-day and future impacts of SO4(2-) and associated trace elements on receiving waters. Synchrotron-based X-ray absorption near edge spectroscopic analysis of coal waste rock samples from the Elk Valley, British Columbia showed Fe present as pyrite (mean 6.0%), siderite (mean 44.3%), goethite (mean 35.4%), and lepidocrocite (mean 14.3%) with S present as sulfide (mean 26.9%), organic S (mean 58.7%), and SO4(2-) (mean 14.4%). Squeezed porewater samples from dump solids yielded mean concentrations of 0.28mg/L Fe and 1246mg/L SO4(2-). Geochemical modeling showed the porewaters in the dumps to be supersaturated with respect to Fe oxyhydroxides and undersaturated with respect to gypsum, consistent with solids analyses. Coupling Fe and S mineralogical data with long-term water quality and quantity measurements from the base of one dump suggest about 10% of the sulfides (which represent 2% of total S) in the dump were oxidized over the past 30years. The S from these oxidized sulfides was released to the receiving surface water as SO4(2-) and the majority of the Fe precipitated as secondary Fe oxyhydroxides (only 3.0×10(-5)% of the Fe was released to the receiving waters over the past 30years). Although the data suggest that the leaching of SO4(2-) from the waste rock dump could continue for about 300years

  3. A novel conditioning process for enhancing dewaterability of waste activated sludge by combination of zero-valent iron and persulfate.

    PubMed

    Zhou, Xu; Wang, Qilin; Jiang, Guangming; Liu, Peng; Yuan, Zhiguo

    2015-06-01

    Improvement of sludge dewaterability is crucial for reducing the costs of sludge disposal in wastewater treatment plants. This study presents a novel conditioning method for improving waste activated sludge dewaterability by combination of persulfate and zero-valent iron. The combination of zero-valent iron (0-30g/L) and persulfate (0-6g/L) under neutral pH substantially enhanced the sludge dewaterability due to the advanced oxidization reactions. The highest enhancement of sludge dewaterability was achieved at 4g persulfate/L and 15g zero-valent iron/L, with which the capillary suction time was reduced by over 50%. The release of soluble chemical oxygen demand during the conditioning process implied the decomposition of sludge structure and microorganisms, which facilitated the improvement of dewaterability due to the release of bound water that was included in sludge structure and microorganism. Economic analysis showed that the proposed conditioning process with persulfate and ZVI is more economically favorable for improving WAS dewaterability than classical Fenton reagent.

  4. Packaging, transportation of LLW

    SciTech Connect

    Shelton, P.

    1994-12-31

    This presentation is an overview of the regulations and requirements for the packaging and transportation of low-level radioactive wastes. United States Environmental Protection Agency and Department of Transportation regulations governing the classification of wastes and the transport documentation are also described.

  5. Remaining Sites Verification Package for the 128-B-2, 100-B Burn Pit #2 Waste Site, Waste Site Reclassification Form 2005-038

    SciTech Connect

    R. A. Carlson

    2005-12-21

    The 128-B-2 waste site was a burn pit historically used for the disposal of combustible and noncombustible wastes, including paint and solvents, office waste, concrete debris, and metallic debris. This site has been remediated by removing approximately 5,627 bank cubic meters of debris, ash, and contaminated soil to the Environmental Restoration Disposal Facility. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  6. Remaining Sites Verification Package for the 600-233 Waste Site, Vertical Pipe Near 100-B Electrical Laydown Area, Waste Site Reclassification Form 2005-041

    SciTech Connect

    R. A. Carlson

    2005-12-08

    The 600-233 waste site consisted of three small-diameter pipelines within the 600-232 waste site, including previously unknown diesel fuel supply lines discovered during site remediation. The 600-233 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  7. Alteration of municipal solid waste incineration bottom ash focusing on the evolution of iron-rich constituents

    SciTech Connect

    Wei Yunmei; Shimaoka, Takayuki; Saffarzadeh, Amirhomayoun; Takahashi, Fumitake

    2011-09-15

    Municipal solid waste incineration (MSWI) bottom ash contains a considerable amount of Fe-rich constituents. The behaviors of these constituents, such as dissolution and precipitation, are quite important as they regulate the distribution of a series of ions between the liquid (percolated fluid) and solid (ash deposit) phases. This paper studied both fresh and weathered MSWI bottom ash from the mineralogical and geochemical viewpoint by utilizing optical microscopy, scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDX), and powder X-ray diffraction. The analysis results revealed that for the fresh bottom ash, iron preferentially existed in the chemical forms of spinel group (mainly Fe{sub 3}O{sub 4}, and a series of Al- or Ti- substituted varieties), metallic inclusions (including Fe-P, Fe-S, Fe-Cu-Pb), hematite (Fe{sub 2}O{sub 3}) and unburned iron pieces. In the 1-20 years weathered bottom ash collected from a landfill site, interconversions among these Fe-rich constituents were identified. Consequently, numerous secondary products were developed, including goethite ({alpha}-FeOOH), lepidocrocite ({gamma}-FeOOH), hematite, magnetite, wustite (FeO), Fe-Si-rich gel phase. Of all these transformation products, hydrous iron oxides were the most common secondary minerals. Quantitative chemical analysis of these secondary products by SEM/EDX disclosed a strong association between the newly formed hydrous iron oxides and heavy metals (e.g. Pb, Zn, Ni, and Cu). The results of this study suggest that the processes of natural weathering and secondary mineralization contribute to reduction of the potential risks of heavy metals to the surrounding environments.

  8. Vegetation cover and long-term conservation of radioactive waste packages: the case study of the CSM waste disposal facility (Manche District, France).

    PubMed

    Petit-Berghem, Yves; Lemperiere, Guy

    2012-03-01

    The CSM is the first French waste disposal facility for radioactive waste. Waste material is buried several meters deep and protected by a multi-layer cover, and equipped with a drainage system. On the surface, the plant cover is a grassland vegetation type. A scientific assessment has been carried out by the Géophen laboratory, University of Caen, in order to better characterize the plant cover (ecological groups and associated soils) and to observe its medium and long term evolution. Field assessments made on 10 plots were complemented by laboratory analyses carried out over a period of 1 year. The results indicate scenarios and alternative solutions which could arise, in order to passively ensure the long-term safety of the waste disposal system. Several proposals for a blanket solution are currently being studied and discussed, under the auspices of international research institutions in order to determine the most appropriate materials for the storage conditions. One proposal is an increased thickness of these materials associated with a geotechnical barrier since it is well adapted to the forest plants which are likely to colonize the site. The current experiments that are carried out will allow to select the best option and could provide feedback for other waste disposal facility sites already being operated in France (CSFMA waste disposal facility, Aube district) or in other countries.

  9. Remaining Sites Verification Package for the 116-C-3, 105-C Chemical Waste Tanks, Waste Site Reclassification Form 2008-002

    SciTech Connect

    L. M. Dittmer

    2008-01-31

    The 116-C-3 waste site consisted of two underground storage tanks designed to receive mixed waste from the 105-C Reactor Metals Examination Facility chemical dejacketing process. Confirmatory evaluation and subsequent characterization of the site determined that the southern tank contained approximately 34,000 L (9,000 gal) of dejacketing wastes, and that the northern tank was unused. In accordance with this evaluation, the verification sampling and modeling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrate that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also show that residual contaminant concentrations are protective of groundwater and the Columbia River.

  10. Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area, Waste Site Reclassification Form 2006-003

    SciTech Connect

    R. A. Carlson

    2006-04-24

    The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  11. Final evaluation report for Westinghouse Hanford Company, WRAP-1,208 liter waste drum, docket 94-35-7A, type A packaging

    SciTech Connect

    Kelly, D.L., Westinghouse Hanford

    1996-06-12

    This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the Westinghouse Hanford Company, Waste Receiving and Processing Facility, Module 1 (WRAP-1) Drum. The WRAP-1 Drum was tested for DOE-HQ in August 1994, by Los Alamos National Laboratory, under docket number 94-35-7A. Additionally, comparison and evaluation of the approved, as-tested packaging configuration was performed by WHC in September 1995. The WRAP-1 Drum was evaluated against the performance of the DOT-17C, 208 1 (55-gal) steel drums tested and evaluated under dockets 89-13-7A/90-18-7A and 94-37-7A.

  12. Remaining Sites Verification Package for the 128-B-3 Burn Pit Site, Waste Site Reclassification Form 2006-058

    SciTech Connect

    L. M. Dittmer

    2006-11-17

    The 128-B-3 waste site is a former burn and disposal site for the 100-B/C Area, located adjacent to the Columbia River. The 128-B-3 waste site has been remediated to meet the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results of sampling at upland areas of the site also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  13. Preliminary thermal/thermomechanical analyses of the Site Characterization Plan's Conceptual Design for a repository containing horizontally emplaced waste packages at the Deaf Smith County site

    SciTech Connect

    Ghantous, N.Y.; Raines, G.E.

    1987-10-01

    This report presents thermal/thermomechanical analyses of the Site Characterization Plan Conceptual Design for horizontal package emplacement at the Deaf Smith County site, Texas. The repository was divided into three geometric regions. Then two-dimensional finite-element models were set up to approximate the three-dimensional nature of each region. Thermal and quasistatic thermomechanical finite-element analyses were performed to evaluate the thermal/thermomechanical responses of the three regions. The exponential-time creep law was used to represent the creep behavior of salt rock. The repository design was evaluated by comparing the thermal/thermomechanical responses obtained for the three regions with interim performance constraints. The preliminary results show that all the performance constraints are met except for those of the waste package. The following factors were considered in interpreting these results: (1) the qualitative description of the analytical responses; (2) the limitations of the analyses; and (3) either the conclusions based on overall evaluation of limitations and analytical results or the conclusions based on the fact that the repository design may be evaluated only after further analyses. Furthermore, a parametric analysis was performed to estimate the effect of material parameters on the predicted thermal/thermomechanical response. 23 refs., 34 figs., 9 tabs.

  14. In-Package Chemistry Abstraction

    SciTech Connect

    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, a 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

  15. Potential of modified iron-rich foundry waste for environmental applications: Fenton reaction and Cr(VI) reduction.

    PubMed

    Oliveira, Patrícia E F; Oliveira, Leandro D; Ardisson, José D; Lago, Rochel M

    2011-10-30

    A magnetic fraction (15%) from a waste of foundry sand (WFS), composed of sand, carbon, bentonite clay and iron (10%) was modified by thermal treatment at 400, 600 and 800°C under inert atmosphere. Mössbauer analyses showed that the thermal treatment increased the amount of Fe(3)O(4) from 25 to 55% by reduction of Fe(2)O(3) and highly dispersed Fe(3+) by the carbon present in the waste. The Fe(3)O(4) caused a significant increase on the activity of two important reactions with application in environmental remediation: the Fenton oxidation of indigo carmine dye with H(2)O(2) and the reduction of Cr(VI) to Cr(III). The magnetic fraction of WFS was also mixed with hematite (Fe(2)O(3)) and thermally treated at 400, 600 and 800°C. This treatment produced large amounts of surface Fe(3)O(4) and increased substantially the rate of Fenton reaction as well as Cr(VI) reduction. This reactivity combined with the presence of carbon (an adsorbent for organic contaminants), bentonite clay (an adsorbent for metallic contaminants) and the granulometry/packing/hydrodynamic features make WFS a promising material for use in reactive permeable barriers.

  16. Remaining Sites Verification Package for the 100-F-46, 119-F Stack Sampling French Drain, Waste Site Reclassification Form 2008-021

    SciTech Connect

    J. M. Capron

    2008-08-08

    The 100-F-46 french drain consisted of a 1.5 to 3 m long, vertically buried, gravel-filled pipe that was approximately 1 m in diameter. Also included in this waste site was a 5 cm cast-iron pipeline that drained condensate from the 119-F Stack Sampling Building into the 100-F-46 french drain. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  17. Remaining Sites Verification Package for the 100-B-23, 100-B/C Area Surface Debris, Waste Site, Waste Site Reclassification Form 2008-027

    SciTech Connect

    J. M. Capron

    2008-06-16

    The 100-B-23, 100-B/C Surface Debris, waste consisted of multiple locations of surface debris and chemical stains that were identified during an Orphan Site Evaluation of the 100-B/C Area. Evaluation of the collected information for the surface debris features yielded four generic waste groupings: asbestos-containing material, lead debris, oil and oil filters, and treated wood. Focused verification sampling was performed concurrently with remediation. Site remediation was accomplished by selective removal of the suspect hazardous items and potentially impacted soils. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  18. Remaining Sites Verification Package for the 1607-F4 Sanitary Sewer System, Waste Site Reclassification Form 2004-131

    SciTech Connect

    L. M. Dittmer

    2007-12-03

    The 1607-F4 waste site is the former location of the sanitary sewer system that serviced the former 115-F Gas Recirculation Building. The system included a septic tank, drain field, and associated pipeline that were in use from 1944 to 1965. The 1607-F4 waste site received unknown amounts of sanitary sewage from the 115-F Gas Recirculation Building and may have potentially contained hazardous and radioactive contamination. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  19. Data Package for Past and Current Groundwater Flow and Contamination beneath Single-Shell Tank Waste Management Areas

    SciTech Connect

    Horton, Duane G.

    2007-03-16

    This appendix summarizes historic and recent groundwater data collected from the uppermost aquifer beneath the 200 East and 200 West Areas. Although the area of interest is the Hanford Site Central Plateau, most of the information discussed in this appendix is at the scale of individual single-shell tank waste management areas. This is because the geologic, and thus the hydraulic, properties and the geochemical properties (i.e., groundwater composition) are different in different parts of the Central Plateau.

  20. Final evaluation & test report for the standard waste box (docket 01-53-7A) type A packaging

    SciTech Connect

    KELLY, D L

    2001-10-15

    This report documents the U.S. Department of Transportation Specification 7A Type A compliance test and evaluation results of the Standard Waste Box. Testing and evaluation activities documented herein are on behalf of the U.S. Department of Energy-Headquarters, Office of Safety, Health and Security (EM-5), Germantown, Maryland. Duratek Federal Services, Inc., Northwest Operations performed an evaluation of the changes as documented herein under Docket 01-53-7A.

  1. General corrosion, irradiation-corrosion, and environmental-mechanical evaluation of nuclear waste package structural barrier materials. Progress report

    SciTech Connect

    Westerman, R.E.; Pitman, S.G.; Nelson, J.L.

    1982-09-01

    Pacific Northwest Laboratory is studying the general corrosion, irradiation-corrosion, and environmentally enhanced crack propagation of five candidate materials in high-temperature aqueous environments simulating those expected in basalt and tuff repositories. The materials include three cast ferrous materials (ductile cast iron and two low-alloy Cr-Mo cast steels) and two titanium alloys, titanium Grade 2 (commercial purity) and Grade 12 (a Ti-Ni-Mo alloy). The general corrosion results are being obtained by autoclave exposure of specimens to slowly replenished simulated ground water flowing upward through a bed of the appropriate crushed rock (basalt or tuff), which is maintained at the desired test temperature (usually 250/sup 0/C). In addition, tests are being performed in deionized water. Metal penetration rates of iron-base alloys are being derived by stripping off the corrosion product film and weighing the specimen after the appropriate exposure time. The corrosion of titanium alloy specimens is being determined by weight gain methods. The irradiation-corrosion studies are similar to the general corrosion tests, except that the specimen-bearing autoclaves are held in a /sup 60/Co gamma radiation field at dose rates up to 2 x 10/sup 6/ rad/h. For evaluating the resistance of the candidate materials to environmentally enhanced crack propagation, three methods are being used: U-bend and fracture toughness specimens exposed in autoclaves; slow strain rate studies in repository-relevant environments to 300/sup 0/C; and fatigue crack growth rate studies at ambient pressure and 90/sup 0/C. The preliminary data suggest a 1-in. corrosion allowance for iron-base barrier elements intended for 1000-yr service in basalt or tuff repositories. No evidence has yet been found that titanium Grade 2 or Grade 12 is susceptible to environmentally induced crack propagation or, by extension, to stress corrosion cracking.

  2. An alternative host matrix based on iron phosphate glasses for the vitrification of specialized nuclear waste forms. 1998 annual progress report

    SciTech Connect

    Day, D.E.; Ray, C.S.; Marasinghe, G.K.; Karabulut, M.; Fang, X.

    1998-06-01

    'Certain high level wastes (HLW) in the US contain components such as phosphates, heavy metals, and halides which make them poorly suited for disposal in borosilicate glasses. Iron phosphate glasses appear to be a technically feasible alternative to borosilicate glasses for vitrifying these HLWs. The iron phosphate glasses mentioned above and their nuclear wasteforms are relatively new, so little is known about their atomic structure, redox equilibria, structure-property relationships, and crystallization products and characteristics. The objective of this research is to gain such information for the binary iron-phosphate glasses as well as iron phosphate wasteforms so that a comprehensive scientific assessment can be made of their usefulness in nuclear waste disposal. This report summarizes the work undertaken and completed in the first 20 months of a three year project. Approximately 250 samples, binary iron phosphate glasses and iron phosphate glasses containing one or two common nuclear waste components such as UO{sub 2} , Na{sub 2}O, Bi{sub 2}O{sub 3} , Cs{sub 2}O, SrO, and MoO{sub 3}, have been prepared. Weight loss has been used to measure the chemical durability and the redox equilibria between Fe(II) and Fe(III) has been investigated using Moessbauer spectroscopy. The atomic structure has been investigated using a variety of techniques including Mossbauer, Raman, X-ray absorption (XAS), and X-ray photoelectron (XPS) spectroscopies and neutron/high energy X-ray scattering. Glass forming and crystallization characteristics have been investigated using differential thermal analysis (DTA). In addition, information necessary for glass manufacturing such as suitable refractories and Joule heating parameters also have been obtained.'

  3. Properties and effects of remaining carbon from waste plastics gasifying on iron scale reduction.

    PubMed

    Zhang, Chongmin; Chen, Shuwen; Miao, Xincheng; Yuan, Hao

    2011-06-01

    The carbonous activities of three kinds of carbon-bearing materials gasified from plastics were tested with coal coke as reference. The results showed that the carbonous activities of these remaining carbon-bearing materials were higher than that of coal-coke. Besides, the fractal analyses showed that the porosities of remaining carbon-bearing materials were higher than that of coal-coke. It revealed that these kinds of remaining carbon-bearing materials are conducive to improve the kinetics conditions of gas-solid phase reaction in iron scale reduction.

  4. Remaining Sites Verification Package for the 100-B-26 Spillway, Waste Site Reclassification Form 2006-052

    SciTech Connect

    L. M. Dittmer

    2006-09-18

    The 100-B-26 Spillway waste site is a spillway that served as an emergency discharge point for the 132-C-2 outfall in the event that the 100-B-15 river effluent pipelines were blocked, damaged, or undergoing maintenance. The selected action involved demonstrating through confirmatory sampling that cleanup goals have been met and proposing a reclassification of this site to No Action. The results of the confirmatory sampling demonstrate that residual contaminant concentrations remaining in the soil beneath the riprap are more protective of groundwater and the Columbia River than the risk they would pose if the site was remediated.

  5. Final evaluation & test report for the standard waste box (docket 01-53-7A) type A packaging

    SciTech Connect

    KELLY, D.L.

    2001-08-15

    This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test and evaluation results of the Standard Waste Box (SWB). Testing and evaluation activities documented herein are on behalf of the U.S. Department of Energy-Headquarters (DOE-HQ), Office of Safety, Health and Security (EM-5), Germantown, Maryland. Dwatek Federal Services, Inc., Northwest Operations (DFSNW) performed an evaluation of the changes as documented herein under Docket 01-53-7A.

  6. Remaining Sites Verification Package for the 1607-F3 Sanitary Sewer System, Waste Site Reclassification Form 2006-047

    SciTech Connect

    L. M. Dittmer

    2007-04-26

    The 1607-F3 waste site is the former location of the sanitary sewer system that supported the 182-F Pump Station, the 183-F Water Treatment Plant, and the 151-F Substation. The sanitary sewer system included a septic tank, drain field, and associated pipeline, all in use between 1944 and 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  7. Remaining Sites Verification Package for 132-D-3, 1608-D Effluent Pumping Station, Waste Site Reclassification Form 2005-033

    SciTech Connect

    R. A. Carlson

    2006-05-09

    Decommissioning and demolition of the 132-D-3 site, 1608-D Effluent Pumping Station was performed in 1986. Decommissioning included removal of equipment, water, and sludge for disposal as radioactive waste. The at- and below-grade structure was demolished to at least 1 m below grade and the resulting rubble buried in situ. The area was backfilled to grade with at least 1 m of clean fill and contoured to the surrounding terrain. Residual concentrations support future land uses that can be represented by a rural-residential scenario and pose no threat to groundwater or the Columbia River based on RESRAD modeling.

  8. Efficiency Assessment of Using Flammable Compounds from Water Treatment and Methanol Production Waste for Plasma Synthesis of Iron-Containing Pigments

    NASA Astrophysics Data System (ADS)

    Shekhovtsova, Anastasia P.; Karengin, Alexander G.

    2016-08-01

    This article describes the possibility of applying the low-temperature plasma for obtaining iron-containing pigments from water purification and flammable methanol production waste. In this paper were calculated combustion parameters of water-saltorganic compositions (WSOC) with different consists. Authors determined the modes of energy- efficient processing of the previously mentioned waste in an air plasma. Having considered the obtained results there were carried out experiments with flammable dispersed water-saltorganic compositions on laboratory plasma stand. All the experimental results are confirmed by calculations.

  9. Three dimensional reconstruction of activity profiles in 220 liters radioactive waste packages containing super-compacted 100 liters drums

    SciTech Connect

    Van Velzen, L.P.M.; Maes, J.

    2007-07-01

    The 3DRedact project's main objective is the development of a non-destructive assay (NDA) system that can replace emission computer tomography (ECT) and transmission computer tomography (TCT) for the routine characterization of decayed radioactive waste 220 liters drums. The existing fast NDA scan system has been extended with a transmission system that fulfils the requirements of fast scan measurements. The design parameters and engineering are described. As a consequence of this extension the analyze program HOLIS had to be updated, so that HOLIS can make full advantage of the transmission data generated by the analysis of a 220 liters waste drum, containing different super compacted drums. The achievements of the new HOLIS version are presented. As a first assessment, based on the presented tests results, the accuracy of the calculated coordinates of hotspots can be assessed for all coordinates {+-} 1 cm and for the activity of the hot-spot {+-} 5 %. These accuracies are within the predefined requirements e.g. coordinates uncertainty {+-} 2 cm and activity less than 10 %. Further, additional safety systems have been installed to improve a healthy and save working environment. (authors)

  10. Remaining Sites Verification Package for the 128-F-2, 100-F Burning Pit Waste Site, Waste Site Reclassification Form 2008-031

    SciTech Connect

    J. M. Capron

    2008-12-01

    The 128-F-2 waste site consisted of multiple burn and debris filled pits located directly east of the 107-F Retention Basin and approximately 30.5 m east of the northeast corner of the 100-F Area perimeter road that runs along the riverbank. The burn pits were used for incinerating nonradioactive, combustible materials from 1945 to 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  11. Interactive effect of trivalent iron on activated sludge digestion and biofilm structure in attached growth reactor of waste tire rubber.

    PubMed

    Sharafat, Iqra; Saeed, Dania Khalid; Yasmin, Sumera; Imran, Asma; Zafar, Zargona; Hameed, Abdul; Ali, Naeem

    2017-03-16

    Waste tire rubber (WTR) has been introduced as an alternative, novel media for biofilm development in several experimental systems including attached growth bioreactors. In this context, four laboratory-scale static batch bioreactors containing WTR as a support material for biofilm development were run under anoxic condition for 90 days using waste activated sludge as an inoculum under the influence of different concentrations (2.5, 6.5, 8.5 mg/l) of trivalent ferric iron (Fe(3+)). The data revealed that activated sludge with a Fe(3+) concentration of 8.5 mg/l supported the maximum bacterial biomass [4.73E + 10 CFU/ml cm(2)]; besides, it removed 38% more Chemical oxygen demand compared to Fe(3+) free condition from the reactor. Biochemical testing and 16S rDNA phylogenetic analysis of WTR-derived biofilm communities further suggested the role of varying concentrations of Fe(3+) on the density and diversity of members of Enterobacteria(ceae), ammonium (AOB) and nitrite oxidizing bacteria. Furthermore, Fluorescent in situ hybridization with phylogenetic oligonucleotide probes and confocal laser scanning microscopy of WTR biofilms indicated a significant increase in density of eubacteria (3.00E + 01 to.05E + 02 cells/cm(2)) and beta proteobacteria (8.10E + 01 to 1.42E + 02 cells/cm(2)), respectively, with an increase in Fe(3+) concentration in the reactors, whereas, the cell density of gamma proteobacteria in biofilms decreased.

  12. Remaining Sites Verification Package for the 126-F-2, 183-F Clearwells, Waste Site Reclassification Form 2006-017

    SciTech Connect

    R. A. Carlson

    2006-05-04

    The 126-F-2 site is the clearwell facility formerly used as part of the reactor cooling water treatment at the 183-F facility. During demolition operations in the 1970s, potentially contaminated debris was disposed in the eastern clearwell structure. The site has been remediated by removing all debris in the clearwell structure to the Environmental Restoration Disposal Facility. The results of radiological surveys and visual inspection of the remediated clearwell structure show neither residual contamination nor the potential for contaminant migration beyond the clearwell boundaries. The results of verification sampling at the remediation waste staging area demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  13. Remaining Sites Verification Package for the 1607-B1 Septic System, Waste Site Reclassification Form 2007-015

    SciTech Connect

    L. M. Dittmer

    2007-08-30

    The 1607-B1 Septic System includes a septic tank, drain field, and associated connecting pipelines and influent sanitary sewer lines. This septic system serviced the former 1701-B Badgehouse, 1720-B Patrol Building/Change Room, and the 1709-B Fire Headquarters. The 1607-B1 waste site received unknown amounts of nonhazardous, nonradioactive sanitary sewage from these facilities during its operational history from 1944 to approximately 1970. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  14. Remaining Sites Verification Package for the 100-F-54 Animal Farm Pastures, Waste Site Reclassification Form 2008-015

    SciTech Connect

    J. M. Capron

    2008-04-17

    The 100-F-54 waste site, part of the 100-FR-2 Operable Unit, is the soil associated with the former pastures for holding domestic farm animals used in experimental toxicology studies. Evaluation of historical information resulted in identification of the experimental animal farm pastures as having potential residual soil contamination due to excrement from experimental animals. The 100-F-54 animal farm pastures confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Studies on solvent extraction of iron(III) as a step for conversion of a waste effluent to a value added product.

    PubMed

    Agrawal, Archana; Kumari, S; Sahu, K K

    2011-12-01

    Solvent extraction of iron(III) from actual sulphate waste pickle liquor was investigated using trialkylphosphine oxide diluted with kerosene. The waste pickle liquor was procured from a local company which deals with the manufacturing of pipes and tubes made of iron and steel. Various parameters were studied to optimise a suitable condition for the maximum extraction of iron. The composition of the aqueous feed used in the experiment was 60.88 g/L Fe(III), 53 g/L acid with traces of Cu, Ni and Co. An ambient extraction at 30 °C yielded acceptable kinetics and loading efficiency for 40% trialkylphosphine oxide with a saturated loading capacity of 51.85 g/L in four contacts at O/A ratio of 1/1 in a multiple contact mode. Iron from the loaded organic was stripped using various strippants such as distilled water, H(2)SO(4) and oxalic acid. Since only 32% of loaded Fe could be stripped with 2 M H(2)SO(4) in five contacts, further stripping was done with 5% oxalic acid which showed a very promising result. It was found that almost 100% of Fe(III) could be stripped out with 5% oxalic acid at O/A of 1/1 in five contacts.

  16. Cost-Effectiveness of Price Subsidies on Fortified Packaged Infant Cereals in Reducing Iron Deficiency Anemia in 6-23-Month-Old-Children in Urban India

    PubMed Central

    Plessow, Rafael; Arora, Narendra Kumar; Brunner, Beatrice

    2016-01-01

    Introduction Iron deficiency anaemia (IDA) is a major public health problem in India and especially harmful in early childhood due to its impact on cognitive development and increased all-cause mortality. We estimate the cost-effectiveness of price subsidies on fortified packaged infant cereals (F-PICs) in reducing IDA in 6-23-monthold children in urban India. Materials and Methods Cost-effectiveness is estimated by comparing the net social cost of price subsidies with the disability-adjusted life-years (DALYs) averted with price subsidies. The net social costs correspond to the cost of the subsidy minus the monetary costs saved by reducing IDA. The estimation proceeds in three steps: 1) the current lifetime costs of IDA are assessed with a health economic model combining the prevalence of anemia, derived from a large population survey, with information on the health consequences of IDA and their costs in terms of mortality, morbidity, and DALYs. 2) The effects of price subsidies on the demand for F-PICs are assessed with a market survey among 4801 households in 12 large Indian cities. 3) The cost-effectiveness is calculated by combining the findings of the first two steps with the results of a systematic review on the effectiveness of F-PICs in reducing IDA. We compare the cost-effectiveness of interventions that differ in the level of the subsidy and in the socio-economic strata (SES) eligible for the subsidy. Results The lifetime social costs of IDA in 6-23-month-old children in large Indian cities amount to production losses of 3222 USD and to 726,000 DALYs. Poor households incur the highest costs, yet even wealthier households suffer substantial losses. The market survey reveals that few households currently buy F-PICs, with the share ranging from 14% to 36%. Wealthier households are generally more likely to buy FPICs. The costs of the subsidies per DALY averted range from 909 to 3649 USD. Interventions targeted at poorer households are most effective. Almost

  17. Improving dewaterability of waste activated sludge by combined conditioning with zero-valent iron and hydrogen peroxide.

    PubMed

    Zhou, Xu; Wang, Qilin; Jiang, Guangming; Zhang, Xiwang; Yuan, Zhiguo

    2014-12-01

    Improvement of sludge dewaterability is crucial for reducing the costs of sludge disposal in wastewater treatment plants. This study presents a novel method based on combined conditioning with zero-valent iron (ZVI) and hydrogen peroxide (HP) at pH 2.0 to improve dewaterability of a full-scale waste activated sludge (WAS). The combination of ZVI (0-750mg/L) and HP (0-750mg/L) at pH 2.0 substantially improved the WAS dewaterability due to Fenton-like reactions. The highest improvement in WAS dewaterability was attained at 500mg ZVI/L and 250mg HP/L, when the capillary suction time of the WAS was reduced by approximately 50%. Particle size distribution indicated that the sludge flocs were decomposed after conditioning. Economic analysis showed that combined conditioning with ZVI and HP was a more economically favorable method for improving WAS dewaterability than the classical Fenton reaction based method initiated by ferrous salts and HP.

  18. Zero-valent iron enhanced methanogenic activity in anaerobic digestion of waste activated sludge after heat and alkali pretreatment.

    PubMed

    Zhang, Yaobin; Feng, Yinghong; Quan, Xie

    2015-04-01

    Heat or alkali pretreatment is the effective method to improve hydrolysis of waste sludge and then enhance anaerobic sludge digestion. However the pretreatment may inactivate the methanogens in the sludge. In the present work, zero-valent iron (ZVI) was used to enhance the methanogenic activity in anaerobic sludge digester under two methanogens-suppressing conditions, i.e. heat-pretreatment and alkali condition respectively. With the addition of ZVI, the lag time of methane production was shortened, and the methane yield increased by 91.5% compared to the control group. The consumption of VFA was accelerated by ZVI, especially for acetate, indicating that the acetoclastic methanogenesis was enhanced. In the alkali-condition experiment, the hydrogen produced decreased from 27.6 to 18.8 mL when increasing the ZVI dosage from 0 to 10 g/L. Correspondingly, the methane yield increased from 1.9 to 32.2 mL, which meant that the H2-utilizing methanogenes was enriched. These results suggested that the addition of ZVI into anaerobic digestion of sludge after pretreated by the heat or alkali process could efficiently recover the methanogenic activity and increase the methane production and sludge reduction.

  19. Inhibiting excessive acidification using zero-valent iron in anaerobic digestion of food waste at high organic load rates.

    PubMed

    Kong, Xin; Wei, Yonghong; Xu, Shuang; Liu, Jianguo; Li, Huan; Liu, Yili; Yu, Shuyao

    2016-07-01

    Excessive acidification occurs frequently in food waste (FW) anaerobic digestion (AD) due to the high carbon-to-nitrogen ratio of FW. In this study, zero-valent iron (ZVI) was applied to prevent the excessive acidification. All of the control groups, without ZVI addition (pH∼5.3), produced little methane (CH4) and had high volatile fatty acids/bicarbonate alkalinity (VFA/ALK). By contrast, at OLR of 42.32gVS/Lreactor, the pH of effluent from the reactors with 0.4g/gVSFWadded of ZVI increased to 7.8-8.2, VFA/ALK decreased to <0.1, and the final CH4 yield was ∼380mL/gVSFWadded, suggesting inhibition of excessive acidification. After adding powdered or scrap metal ZVI to the acidogenic reactors, the fractional content of butyric acid changed from 30-40% to 0%, while, that of acetic acid increased. These results indicate that adding ZVI to FW digestion at high OLRs could eliminate excessive acidification by promoting butyric acid conversion and enhancing methanogen activity.

  20. The removal of iron from molten aluminium

    SciTech Connect

    Donk, H.M. van der; Nijhof, G.H.; Castelijns, C.A.M.

    1995-12-31

    In this work an overview is given about the techniques available for the removal of metallic impurities from molten aluminium. The overview is focused on the removal of iron. Also, some experimental results are given about the creation of iron-rich intermetallic compounds in an aluminium system, which are subsequently removed by gravity segregation and filtration techniques. This work is part of an ongoing research project of three major European aluminium companies who are co-operating on the subject of recycling of aluminium packaging materials recovered from household waste by means of Eddy-Current techniques. Using this technique the pick-up of some contaminating metals, particularly iron, is almost unavoidable.

  1. Remaining Sites Verification Package for the 100-F-50 Stormwater Runoff Culvert, Waste Site Reclassification Form 2007-001

    SciTech Connect

    J. M. Capron

    2008-04-15

    The 100-F-50 waste site, part of the 100-FR-2 Operable Unit, is a steel stormwater runoff culvert that runs between two railroad grades in the south-central portion of the 100-F Area. The culvert exiting the west side of the railroad grade is mostly encased in concrete and surrounded by a concrete stormwater collection depression partially filled with soil and vegetation. The drain pipe exiting the east side of the railroad grade embankment is partially filled with soil and rocks. The 100-F-50 stormwater diversion culvert confirmatory sampling results support a reclassification of this site to no action. The current site conditions achieve the remedial action objectives and corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  2. Numerical studies of fluid and heat flow near high-level nuclear waste packages emplaced in partially saturated fractured tuff

    SciTech Connect

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

    1984-11-01

    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 repository horizon 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 "TOUGH" used for our flow calculations takes into account most of the physical effects which are important in multi-phase fluid and heat flow. It has provisions for handling the extreme non-linearities which 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. From an analysis of the results obtained with explicit fractures, we develop equivalent continuum models which can reproduce the temperature, saturation, and pressure variation, and gas and liquid flow rates of the discrete fracture-porous matrix calculations. The equivalent continuum approach makes use of a generalized relative permeability concept to take into account the fracture effects. This results in a substantial simplification of the flow problem which makes larger scale modeling of complicated unsaturated fractured porous systems feasible. Potential applications for regional scale simulations and limitations of the continuum approach are discussed. 35 refs., 14 figs., 4 tabs.

  3. Immobilization of chromate in hyperalkaline waste streams by green rusts and zero-valent iron.

    PubMed

    Rogers, Christine M; Burke, Ian T; Ahmed, Imad A M; Shaw, Samuel

    2014-01-01

    Zero-valent iron (ZVI) and green rusts can be used as reductants to convert chromium from soluble, highly toxic Cr(VI) to insoluble Cr(III). This study compared the reduction rates of Cr(VI) by ZVI and two carbonate green rust phases in alkaline/hyperalkaline solutions. Batch experiments were carried out with synthetic chromate solutions at pH 7.7-12.3 and a chromite ore processing residue (COPR) leachate (pH approximately 12.2). Green rust removes chromate from high pH solutions (pH 10-12.5) very rapidly (<400 s). Chromate reduction rates for both green rust phases were consistently higher than for ZVI throughout the pH range studied; the surface area normalized rate constants were two orders of magnitude higher in the COPR leachate solution at pH 12.2. The performances of both green rusts were unaffected by changes in pH. In contrast, ZVI exhibited a marked decline in reduction rate with increasing pH to become almost ineffective above pH12.

  4. EFFECTIVENESS OF USING DILUTE OXALIC ACID TO DISSOLVEHIGH LEVEL WASTE IRON BASED SLUDGE SIMULANT

    SciTech Connect

    Ketusky, E

    2008-07-11

    At the Savannah River Site (SRS), near Aiken South Carolina, there is a crucial need to remove residual quantities of highly radioactive iron-based sludge from large select underground storage tanks (e.g., 19,000 liters of sludge per tank), in order to support tank closure. The use of oxalic acid is planned to dissolve the residual sludge, hence, helping in the removal. Based on rigorous testing, primarily using 4 and 8 wt% oxalic acid solutions, it was concluded that the more concentrated the acid, the greater the amount of residual sludge that would be dissolved; hence, a baseline technology on using 8 wt% oxalic acid was developed. In stark contrast to the baseline technology, reports from other industries suggest that the dissolution will most effectively occur at 1 wt% oxalic acid (i.e., maintaining the pH near 2). The driver for using less oxalic acid is that less (i.e., moles) would decrease the severity of the downstream impacts (i.e., required oxalate solids removal efforts). To determine the initial feasibility of using 1 wt% acid to dissolve > 90% of the sludge solids, about 19,000 liters of representative sludge was modeled using about 530,000 liters of 0 to 8 wt% oxalic acid solutions. With the chemical thermodynamic equilibrium based software results showing that 1 wt% oxalic acid could theoretically work, simulant dissolution testing was initiated. For the dissolution testing, existing simulant was obtained, and an approximate 20 liter test rig was built. Multiple batch dissolutions of both wet and air-dried simulant were performed. Overall, the testing showed that dilute oxalic acid dissolved a greater fraction of the stimulant and resulted in a significantly larger acid effectiveness (i.e., grams of sludge dissolved/mole of acid) than the baseline technology. With the potential effectiveness confirmed via simulant testing, additional testing, including radioactive sludge testing, is planned.

  5. Optimal segmentation and packaging process

    DOEpatents

    Kostelnik, Kevin M.; Meservey, Richard H.; Landon, Mark D.

    1999-01-01

    A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D&D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded.

  6. 49 CFR 173.63 - Packaging exceptions.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... waste, marine pollutant, or is offered for transportation and transported byaircraft or vessel... SHIPMENTS AND PACKAGINGS Definitions, Classification and Packaging for Class 1 § 173.63 Packaging exceptions...-shore supply vessel; (3) Cargo compartment of a cargo vessel; or (4) Passenger-carrying aircraft used...

  7. 49 CFR 173.63 - Packaging exceptions.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... SHIPMENTS AND PACKAGINGS Definitions, Classification and Packaging for Class 1 § 173.63 Packaging exceptions...-shore supply vessel; (3) Cargo compartment of a cargo vessel; or (4) Passenger-carrying aircraft used to... hazardous substance, hazardous waste, marine pollutant, or is offered for transportation and transported...

  8. Recovery and separation of sulfuric acid and iron from dilute acidic sulfate effluent and waste sulfuric acid by solvent extraction and stripping.

    PubMed

    Qifeng, Wei; Xiulian, Ren; Jingjing, Guo; Yongxing, Chen

    2016-03-05

    The recovery and simultaneous separation of sulfuric acid and iron from dilute acidic sulfate effluent (DASE) and waste sulfuric acid (WSA) have been an earnest wish for researchers and the entire sulfate process-based titanium pigment industry. To reduce the pollution of the waste acid and make a comprehensive use of the iron and sulfuric acid in it, a new environmentally friendly recovery and separation process for the DASE and the WSA is proposed. This process is based on the reactive extraction of sulfuric acid and Fe(III) from the DASE. Simultaneously, stripping of Fe(III) is carried out in the loaded organic phase with the WSA. Compared to the conventional ways, this innovative method allows the effective extraction of sulfuric acid and iron from the DASE, and the stripping of Fe(III) from the loaded organic phase with the WSA. Trioctylamine (TOA) and tributyl phosphate (TBP) in kerosene (10-50%) were used as organic phases for solvent extraction. Under the optimal conditions, about 98% of Fe(III) and sulfuric acid were removed from the DASE, and about 99.9% of Fe(III) in the organic phase was stripped with the WSA.

  9. DROP TESTS RESULTS OF REVISED CLOSURE BOLT CONFIGURATION OF THE STANDARD WASTE BOX, STANDARD LARGE BOX 2, AND TEN DRUM OVERPACK PACKAGINGS

    SciTech Connect

    May, C.; Opperman, E.; Mckeel, C.

    2010-04-15

    The Transuranic (TRU) Disposition Project at Savannah River Site will require numerous transfers of radioactive materials within the site boundaries for sorting and repackaging. The three DOT Type A shipping packagings planned for this work have numerous bolts for securing the lids to the body of the packagings. In an effort to reduce operator time to open and close the packages during onsite transfers, thus reducing personnel exposure and costs, an evaluation was performed to analyze the effects of reducing the number of bolts required to secure the lid to the packaging body. The evaluation showed the reduction to one-third of the original number of bolts had no effect on the packagings capability to sustain vibratory loads, shipping loads, internal pressure loads, and the loads resulting from a 4-ft drop. However, the loads caused by the 4-ft drop are difficult to estimate and the study recommended each of the packages be dropped to show the actual effects on the package closure. Even with reduced bolting, the packagings were still required to meet the 49 CFR 178.350 performance criteria for Type A packaging. This paper discusses the effects and results of the drop testing of the three packagings.

  10. Recovery of polypropylene and polyethylene from packaging plastic wastes without contamination of chlorinated plastic films by the combination process of wet gravity separation and ozonation.

    PubMed

    Reddy, Mallampati Srinivasa; Okuda, Tetsuji; Nakai, Satoshi; Nishijima, Wataru; Okada, Mitsumasa

    2011-08-01

    Wet gravity separation technique has been regularly practiced to separate the polypropylene (PP) and polyethylene (PE) (light plastic films) from chlorinated plastic films (CP films) (heavy plastic films). The CP films including poly vinyl chloride (PVC) and poly vinylidene chloride (PVDC) would float in water even though its density is more than 1.0g/cm(3). This is because films are twisted in which air is sometimes entrapped inside the twisted CP films in real existing recycling plant. The present research improves the current process in separating the PP and PE from plastic packaging waste (PPW), by reducing entrapped air and by increasing the hydrophilicity of the CP films surface with ozonation. The present research also measures the hydrophilicity of the CP films. In ozonation process mixing of artificial films up to 10min reduces the contact angle from 78° to 62°, and also increases the hydrophilicity of CP films. The previous studies also performed show that the artificial PVDC films easily settle down by the same. The effect of ozonation after the wet gravity separation on light PPW films obtained from an actual PPW recycling plant was also evaluated. Although actual light PPW films contained 1.3% of CP films however in present case all the CP films were removed from the PPW films as a settled fraction in the combination process of ozonation and wet gravity separation. The combination process of ozonation and wet gravity separation is the more beneficial process in recovering of high purity PP and PE films from the PPW films.

  11. MMIC packaging with Waffleline

    NASA Astrophysics Data System (ADS)

    Perry, R. W.; Ellis, T. T.; Schineller, E. R.

    1990-06-01

    The design principle of Waffleline, a patented MMIC packaging technology, is discussed, and several recent applications are described and illustrated with drawings, diagrams, and photographs. Standard Waffleline is a foil-covered waffle-iron-like grid with dielectric-coated signal and power wires running in the channels and foil-removed holes for mounting prepackaged chips or chip carriers. With spacing of 50 mils between center conductors, this material is applicable at frequencies up to 40 GHz; EHF devices require Waffleline with 25-mil spacing. Applications characterized include a subassembly for a man-transportable SHF satellite-communication terminal, a transmitter driver for a high-power TWT, and a 60-GHz receiver front end (including an integrated monolithic microstrip antenna, a low-noise amplifier, a mixer, and an IF amplifier in a 0.25-inch-thick 1.6-inch-diameter package). The high package density and relatively low cost of Waffleline are emphasized.

  12. RH Packaging Program Guidance

    SciTech Connect

    Washington TRU Solutions LLC

    2008-01-12

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the "RH-TRU 72-B cask") and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a

  13. Feeding practices and factors contributing to wasting, stunting, and iron-deficiency anaemia among 3-23-month old children in Kilosa district, rural Tanzania.

    PubMed

    Mamiro, Peter S; Kolsteren, Patrick; Roberfroid, Dominique; Tatala, Simon; Opsomer, Ann S; Van Camp, John H

    2005-09-01

    Infants in Tanzania are particularly vulnerable to under-nutrition during transition from breastmilk (as the only source of nourishment) to solid foods. A cross-sectional study was undertaken in Kilosa district in Tanzania to determine the feeding practices and the extent of wasting, stunting, and iron-deficiency anaemia. The study was done in two stages: in the first stage, a 24-hour dietary assessment was conducted to identify the type of complementary foods given and the eating habits according to age for 378 children aged 3-23 months. In the second stage, a progressive recruitment of 309 infants aged six months was made to measure weight, length, haemoglobin (Hb) concentration, zinc protoporphyrin concentration, and malaria parasitaemia. Birth-weight, the potential contributing factor to under-nutrition and iron-deficiency anaemia, was obtained from the children's clinic cards. The 24-hour dietary assessment revealed that children consumed mainly a thin porridge prepared from maize flour as complementary food. Carbohydrates contributed most energy (on average 69%), followed by fats (18.6%) and protein (on average 12.1%). The complementary food co-vered only 15%, 20%, and 27% of the recommended iron intake for children aged 6-8, 9-11 and 12-23 months respectively. The mean Hb concentration was 9.3 +/- 1.9 g/dL, 68% of the infants were moderately anaemic (7 < or =11 g/dL), and about 11% were severely anaemic with Hb below 7 g/dL, while 21% were non-anaemic Hb (> or =11 g/dL). Equally, the mean zinc protoporphyrin concentration was 10.0 +/- 6.2 microg/g Hb, and 76% of the infants were iron-deficient (>5 microg/g Hb). The prevalence of stunting was 35%, while wasting was only 1.3%. Low birth-weight and low body mass index of mothers were the strong predictors of stunting, whereas low birth-weight and iron-deficiency were the strong predictors of anaemia. The prevalence of malaria parasitaemia was high, affecting 50% of the infants. Having malaria was the only

  14. Immobilization of Iron Nanoparticles on Multi Substrates and Its Reduction Removal of Chromium (VI) from Waste Streams

    EPA Science Inventory

    This article describes the in-situ synthesis and immobilization of iron nanoparticles on several substrates at room temperature using NaBH4 as a reducing agent and ascorbic acid as capping agent. The method is very effective in protecting iron nanoparticles from air oxidation for...

  15. Assessing materials (''Getters'') to immobilize or retard the transport of technetium through the engineered barrier system at the potential Yucca Mountain nuclear waste repository

    SciTech Connect

    Viani, B E

    1999-03-15

    Current performance assessment calculations show that technetium (Tc) and neptunium (Np) will deliver the major fraction of the radiation dose to the accessible environment from the potential Yucca Mountain nuclear waste repository. Therefore, materials that can immobilize or delay the transport of Tc or Np (getters) are being considered for addition to either the waste-package or the backfill adjacent to the waste-package. Of the two radionuclides, Tc presents the greater challenge in identifying a suitable getter material. This report identifies several materials that warrant further consideration for immobilizing and/or sorbing Tc as additives to the backfill, and recommends active carbon and an inorganic oxide for initial testing. Other materials, such as zero valent iron, might be useful as getters if they were placed in the waste package itself, a subject that merits further investigation.

  16. Scoring Package

    National Institute of Standards and Technology Data Gateway

    NIST Scoring Package (PC database for purchase)   The NIST Scoring Package (Special Database 1) is a reference implementation of the draft Standard Method for Evaluating the Performance of Systems Intended to Recognize Hand-printed Characters from Image Data Scanned from Forms.

  17. CH Packaging Program Guidance

    SciTech Connect

    Washington TRU Solutions LLC

    2005-02-28

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

  18. Food Packaging Materials

    NASA Technical Reports Server (NTRS)

    1978-01-01

    The photos show a few of the food products packaged in Alure, a metallized plastic material developed and manufactured by St. Regis Paper Company's Flexible Packaging Division, Dallas, Texas. The material incorporates a metallized film originally developed for space applications. Among the suppliers of the film to St. Regis is King-Seeley Thermos Company, Winchester, Ma'ssachusetts. Initially used by NASA as a signal-bouncing reflective coating for the Echo 1 communications satellite, the film was developed by a company later absorbed by King-Seeley. The metallized film was also used as insulating material for components of a number of other spacecraft. St. Regis developed Alure to meet a multiple packaging material need: good eye appeal, product protection for long periods and the ability to be used successfully on a wide variety of food packaging equipment. When the cost of aluminum foil skyrocketed, packagers sought substitute metallized materials but experiments with a number of them uncovered problems; some were too expensive, some did not adequately protect the product, some were difficult for the machinery to handle. Alure offers a solution. St. Regis created Alure by sandwiching the metallized film between layers of plastics. The resulting laminated metallized material has the superior eye appeal of foil but is less expensive and more easily machined. Alure effectively blocks out light, moisture and oxygen and therefore gives the packaged food long shelf life. A major packaging firm conducted its own tests of the material and confirmed the advantages of machinability and shelf life, adding that it runs faster on machines than materials used in the past and it decreases product waste; the net effect is increased productivity.

  19. Use of iron and bio-oil wastes to produce highly dispersed Fe/C composites for the photo-Fenton reaction.

    PubMed

    de Mendonça, Fernanda Gomes; Rosmaninho, Marcelo Gonçalves; da Fonseca, Philipe Xavier; Soares, Ricardo Reis; Ardisson, José Domingos; Tristão, Juliana Cristina; Lago, Rochel Montero

    2017-03-01

    This work describes the synthesis, characterization, and application of an active heterogeneous photo-Fenton system obtained from two different wastes, i.e., laterite (an iron mining waste) and the acid aqueous fraction (AAF) from bio-oil production. AAF with high acidity (ca. 3 molH+ L(-1)) and organic concentration (25 wt.%) obtained from biomass flash pyrolysis was used for the efficient extraction of Fe(3+) from laterite waste. After extraction, the mixture Fe(3+)/AAF was dried and treated at different temperatures, i.e., 500, 650, and 800 °C, to obtain Fe/C reactive composites. Mössbauer, XRD, TG, elemental analyses, and SEM/EDS showed the presence of highly disperse Fe oxide nanoparticles at 500 and 650 °C and Fe(0) particles in the material obtained at 800 °C with carbon contents varying from 74 to 80 %. The three composites were tested as heterogeneous catalysts in the photo-Fenton reaction for the oxidation of the model dye contaminant methylene blue, showing high activities at neutral pH.

  20. Teaching Old Packaging New Tricks - 12593

    SciTech Connect

    England, Jeffery L.; Shuler, James M.

    2012-07-01

    Waste disposition campaigns have been an industry and government focus area since the mid- 1970's. With increased focus on this issue, and a lot of hard work, most waste packaging and transportation issues have been addressed. The material has been successfully shipped and dis-positioned. DOE has successfully de-inventoried materials from multiple sites to meet material consolidation, footprint reduction, nonproliferation, and regulatory obligations with cost savings from reduced maintenance and regulatory compliance. There has been a wide range of certified shipping packagings for the transportation of hazardous materials to meet most of the waste needs. The remaining materials are problematic, generally low volume, and do not meet the certified content of the existing inventory of packaging. Designing, testing and certifying new packaging designs can be a long and expensive process and for small volumes of material it is cost prohibitive. One very cost effective option is to lease and use a certified packaging to overpack waste containers. There are many robust certified packagings available with the capability to envelope the waste content. The capability to use inner containers, inside the current fleet of certified casks or packaging, to address specific content problems of additional shielding (e.g., U-233) or containment (e.g., sodium bonded nuclear material) has successfully expanded the capability for timely cost effective shipment of unique contents. This option has been used successfully in the NAC-LWT, T-3 and other packagings. (authors)

  1. Immobilization of antimony in waste-to-energy bottom ash by addition of calcium and iron containing additives.

    PubMed

    Van Caneghem, Jo; Verbinnen, Bram; Cornelis, Geert; de Wijs, Joost; Mulder, Rob; Billen, Pieter; Vandecasteele, Carlo

    2016-08-01

    The leaching of Sb from waste-to-energy (WtE) bottom ash (BA) often exceeds the Dutch limit value of 0.32mgkg(-1) for recycling of BA in open construction applications. From the immobilization mechanisms described in the literature, it could be concluded that both Ca and Fe play an important role in the immobilization of Sb in WtE BA. Therefore, Ca and Fe containing compounds were added to the samples of the sand fraction of WtE BA, which in contrast to the granulate fraction is not recyclable to date, and the effect on the Sb leaching was studied by means of batch leaching tests. Results showed that addition of 0.5 and 2.5% CaO, 5% CaCl2, 2.5% Fe2(SO4)3 and 1% FeCl3 decreased the Sb leaching from 0.62±0.02mgkgDM(-1) to 0.20±0.02, 0.083±0.044, 0.25±0.01, 0.27±0.002 and 0.29±0.02mgkgDM(-1), respectively. Due to the increase in pH from 11.41 to 12.53 when 2.5% CaO was added, Pb and Zn leaching increased and exceeded the respective leaching limits. Addition of 5% CaCO3 had almost no effect on the Sb leaching, as evidenced by the resulting 0.53mgkgDM(-1) leaching concentration. This paper shows a complementary enhancement of the effect of Ca and Fe, by comparing the aforementioned Sb leaching results with those of WtE BA with combined addition of 2.5% CaO or 5% CaCl2 with 2.5% Fe2(SO4)3 or 1% FeCl3. These lab scale results suggest that formation of romeites with a high Ca content and formation of iron antimonate (tripuhyite) with a very low solubility are the main immobilization mechanisms of Sb in WtE BA. Besides the pure compounds and their mixtures, also addition of 10% of two Ca and Fe containing residues of the steel industry, hereafter referred to as R1 and R2, was effective in decreasing the Sb leaching from WtE BA below the Dutch limit value for reuse in open construction applications. To evaluate the long term effect of the additives, pilot plots of WtE BA with 10% of R1 and 5% and 10% of R2 were built and samples were submitted to leaching tests at

  2. CH Packaging Program Guidance

    SciTech Connect

    None, None

    2008-09-11

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the pplication." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  3. CH Packaging Program Guidance

    SciTech Connect

    None, None

    2009-06-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  4. Nanoscale Zero-Valent Iron (NZVI) supported on sineguelas waste for Pb(II) removal from aqueous solution: kinetics, thermodynamic and mechanism.

    PubMed

    Arshadi, M; Soleymanzadeh, M; Salvacion, J W L; SalimiVahid, F

    2014-07-15

    In this study, the synthesis and characterization of a new adsorbent containing nanoscale zerovalent iron particles (NZVI) decorated sineguelas waste (S-NaOH-NZVI) from agriculture biomass was investigated for the adsorption/reduction of inorganic pollution such as Pb(II) ions. The combination of ZVI particles on the surface of sineguelas waste can help to overcome the disadvantage of ultra-fine powders which may have strong tendency to agglomerate into larger particles, resulting in an adverse effect on both effective surface area and catalyst performance. The synthesized materials were characterized with different methods such as FT-IR, BET, XRD, TEM and pHPZC. Good dispersion of NZVI particles (ca. 10-70nm) on the sineguelas waste was observed. The effects of various parameters, such as contact time, pH, concentration, adsorbent dosage and temperature were studied. The adsorption of Pb(II) ions has been studied in terms of pseudo-first- and second-order kinetics, and the Freundlich, Langmuir and Langmuir-Freundlich isotherms models have also been used to the equilibrium adsorption data. The adsorption kinetics followed the mechanism of the pseudo-second-order equation. The thermodynamic parameters (ΔG, ΔH and ΔS) indicated that the adsorption of Pb(II) ions were feasible, spontaneous and endothermic at 25-80°C. XRD analysis indicated the presence of Pb(0) on the S-NaOH-NZVI surface. This study suggests that the modified sineguelas waste by NZVI particles can be prepared at low cost and the materials are environmentally benign for the removal of Pb(II) ions, and likely many other heavy metal ions, from water.

  5. CH Packaging Program Guidance

    SciTech Connect

    None, None

    2006-04-25

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package TransporterModel II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) §71.8. Any time a user suspects or has indications that the conditions ofapproval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  6. CH Packaging Program Guidance

    SciTech Connect

    None, None

    2007-12-13

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  7. Remaining Sites Verification Package for the 100-F-31, 144-F Sanitary Sewer System, Waste Site Reclassification Form 2006-033

    SciTech Connect

    L. M. Dittmer

    2006-08-24

    The 100-F-31 waste site is a former septic system that supported the inhalation laboratories, also referred to as the 144-F Particle Exposure Laboratory (132-F-2 waste site), which housed animals exposed to particulate material. The 100-F-31 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  8. 77 FR 60056 - Shippers-General Requirements for Shipments and Packagings

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-02

    ... Shipments and Packagings CFR Correction In Title 49 of the Code of Federal Regulations, Parts 100 to 177... waste when packaged in a rigid non- bulk packaging conforming to the general packaging requirements of Sec. Sec. 173.24 and 173.24a and packaging requirements specified in 29 CFR 1910.1030 and...

  9. Remaining Sites Verification Package for the 116-F-16, PNL Outfall and the 100-F-43, PNL Outfall Spillway, Waste Site Reclassification Form 2006-039

    SciTech Connect

    L. M. Dittmer

    2006-09-14

    The 116-F-16 waste site is the former Pacific National Laboratories (PNL) Outfall, used to discharge waste effluents from the 100-F Experimental Animal Farm. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  10. Cleanup Verification Package for the 618-2 Burial Ground

    SciTech Connect

    W. S. Thompson

    2006-12-28

    This cleanup verification package documents completion of remedial action for the 618-2 Burial Ground, also referred to as Solid Waste Burial Ground No. 2; Burial Ground No. 2; 318-2; and Dry Waste Burial Site No. 2. This waste site was used primarily for the disposal of contaminated equipment, materials and laboratory waste from the 300 Area Facilities.

  11. Iron Cycling in Low pH Environments - Potential Application for the Recovery of Precious Metals from Industrial Waste

    NASA Astrophysics Data System (ADS)

    Muehe, E. M.; Helle, T.; Kappler, A.

    2014-12-01

    The use of many different precious metals (gold, platinum…) and Rare Earth Elements (lanthanum, neodymium…) in the production of electronic products is drastically increasing. To meet this demand, not only mining activities but recently also the recovery of these elements from industrial waste is in the focus of research. It has been shown that the application of extracting solutions with pH values lower than 4 lead to an economically feasible recovery of industrially precious metals. This abiotic extraction efficiency can potentially be increased by using microorganisms capable of dissolving more stable minerals at low pH. In collaboration with industry, a waste incineration plant, and governmental authorities, we investigate the extraction and recovery of strategically important metals and Rare Earth Elements from industrial waste. We optimize the (bio)-geochemical conditions for the extraction of these elements. To this end, a variety of microorganisms are evaluated for efficient metal extraction. We focus on known laboratory cultures capable of oxidizing and reducing Fe minerals and S compounds. Additionally, unknown microbial communities able to increase the efficiency of precious metal extraction from the industrial waste are enriched from environments with comparable geochemical conditions found in the extraction solutions.

  12. High-rate iron-rich activated sludge as stabilizing agent for the anaerobic digestion of kitchen waste.

    PubMed

    De Vrieze, Jo; De Lathouwer, Lars; Verstraete, Willy; Boon, Nico

    2013-07-01

    Anaerobic digestion is a key technology in the bio-based economy and can be applied to convert a wide range of organic substrates into CH4 and CO2. Kitchen waste is a valuable substrate for anaerobic digestion, since it is an abundant source of organic matter. Yet, digestion of single kitchen waste often results in process failure. High-rate activated sludge or A-sludge is produced during the highly loaded first stage of the two-phase 'Adsorptions-Belebungsverfahren' or A/B activated sludge system for municipal wastewater treatment. In this specific case, the A-sludge was amended with FeSO4 to enhance phosphorous removal and coagulation during the water treatment step. This study therefore evaluated whether this Fe-rich A-sludge could be used to obtain stable methanation and higher methane production values during co-digestion with kitchen waste. It was revealed that Fe-rich A-sludge can be a suitable co-substrate for kitchen waste; i.e. methane production rate values of 1.15 ± 0.22 and 1.12 ± 0.28 L L(-1) d(-1) were obtained during mesophilic and thermophilic co-digestion respectively of a feed-mixture consisting of 15% KW and 85% A-sludge. The thermophilic process led to higher residual VFA concentrations, up to 2070 mg COD L(-1), and can therefore be considered less stable. Addition of micro- and macronutrients provided a more stable digestion of single kitchen waste, i.e. a methane production of 0.45 L L(-1) d(-1) was obtained in the micronutrient treatment compared to 0.30 L L(-1) d(-1) in the control treatment on day 61. Yet, methane production during single kitchen waste digestion still decreased toward the end of the experiment, despite the addition of micronutrients. Methane production rates were clearly influenced by the total numbers of archaea in the different reactors. This study showed that Fe-rich A-sludge and kitchen waste are suitable for co-digestion.

  13. Spack: the Supercomputing Package Manager

    SciTech Connect

    Gamblin, T.

    2013-11-09

    The HPC software ecosystem is growing larger and more complex, but software distribution mechanisms have not kept up with this trend. Tools, Libraries, and applications need to run on multiple platforms and build with multiple compliers. Increasingly, packages leverage common software components, and building any one component requires building all of its dependencies. In HPC environments, ABI-incompatible interfaces (likeMPI), binary-incompatible compilers, and cross-compiled environments converge to make the build process a combinatoric nightmare. This obstacle deters many users from adopting useful tools, and others waste countless hours building and rebuilding tools. Many package managers exist to solve these problems for typical desktop environments, but none suits the unique needs of supercomputing facilities or users. To address these problems, we have Spack, a package manager that eases the task of managing software for end-users, across multiple platforms, package versions, compilers, and ABI incompatibilities.

  14. The role of zero valent iron on the fate of tetracycline resistance genes and class 1 integrons during thermophilic anaerobic co-digestion of waste sludge and kitchen waste.

    PubMed

    Gao, Pin; Gu, Chaochao; Wei, Xin; Li, Xiang; Chen, Hong; Jia, Hanzhong; Liu, Zhenhong; Xue, Gang; Ma, Chunyan

    2017-03-15

    Activated sludge has been identified as a potential significant source of antibiotic resistance genes (ARGs) to the environment. Anaerobic digestion is extensively used for sludge stabilization and resource recovery, and represents a crucial process for controlling the dissemination of ARGs prior to land application of digested sludge. The objective of this study is to investigate the effect of zero valent iron (Fe(0)) on the attenuation of seven representative tetracycline resistance genes (tet, tet(A), tet(C), tet(G), tet(M), tet(O), tet(W), and tet(X)), and the integrase gene intI1 during thermophilic anaerobic co-digestion of waste sludge and kitchen waste. Significant decrease (P < 0.05) in the quantities of tet (except tet(W)) and intI1 genes was observed at Fe(0) dosage of 5 g/L, whereas no significant differences (P > 0.05) were found for all gene targets between digesters with Fe(0) dosages of 5 and 60 g/L. A first-order kinetic model favorably described the trends in concentrations of tet and intI1 gene targets during thermophilic anaerobic digestion with or without Fe(0). Notably, tet genes encoding different resistance mechanisms behaved distinctly in anaerobic digesters, although addition of Fe(0) could enhance their reduction. The overall results of this research suggest that thermophilic anaerobic digestion with Fe(0) can be a potential alternative technology for the attenuation of tet and intI1 genes in waste sludge.

  15. Key parameters for low-grade fine-grained iron ore valorization: lower environmental impact through reduced waste.

    NASA Astrophysics Data System (ADS)

    Wagner, Christiane; Orberger, Beate; Tudryn, Alina; Baptiste, Benoît; Wirth, Richard; Morgan, Rachel; Miska, Serge

    2016-04-01

    In low-grade banded iron formations (BIFs), a large part of the iron is related to micro- and nano- metric iron-bearing inclusions within quartz and/or carbonates, mainly dolomite (~ 20 to 50 μm). Low-grade fine grained iron ore present two types of environmental risks: a) they are often stocked as tailings. For example, the recent disaster (5th of November 2015) in the Minas Gerais district, Brazil, was caused by the collapse of the Fundão tailings dam at an open cast mine; b) during beneficiation significant amounts of dust are generated also leading to metal loss. A laminated BIF studied from a drill core at Àguas Claras Mine, Quadrilátero Ferrífero, Brazil, contains 26.71 wt. % total iron, 0.2 wt. % SiO2, 0.32 wt.% MnO, 15.46 wt. % MgO, 22.32 wt.% CaO, 0.09 wt. % P2O5, < 0.05 wt. % Al2O3, 0.15 wt. % H2O and 34.08 wt. % CO2. Environmental hazardous elements are present as traces (As: 3-20 ppm, Cd: 0-0.7 ppm; Cr: 0.05-60 ppm, Pb: up to 55 ppm; U: up to 8 ppm). Dolomite and quartz bands alternate with hematite bands. Raman spectroscopy, X-ray diffraction and FIB-TEM analyses reveal that the micro- and nano- metric inclusions in dolomite are hematite and minor goethite, partly occurring as clusters in voids. Curie Balance analyses were carried out at different heating steps and temperatures on whole rock samples and a synthetic mix of decarbonated sample and pure dolomite. X-ray diffraction on the products of the heating experiments shows that that hematite is stable and new phases: magnesioferrite (MgFe2O4), lime (CaO), periclase (MgO), portlandite (Ca(OH)2) and srebrodoskite (Ca2Fe2O5) were formed between 680 °C and 920 °C. These findings promote the economic use of low grade ores rather than their stockpiling as tailings. The presence of OH-bearing goethite reduces the sintering temperature. After having separated coarse hematite from barren dolomite and quartz, a low temperature sintering of the inclusion-bearing dolomite/quartz leads to transformations

  16. Geochemical Characterization Data Package for the Vadose Zone in the Single-Shell Tank Waste Management Areas at the Hanford Site

    SciTech Connect

    Cantrell, Kirk J.; Brown, Christopher F.; Serne, R. Jeffrey; Krupka, Kenneth M.

    2008-01-07

    This data package discusses the geochemistry of vadose zone sediments beneath the single-shell tank (SST) farms at the U.S. Department of Energy’s (DOE’s) Hanford Site. The purpose of the report is to provide a review of the most recent and relevant geochemical information available for the vadose zone beneath the SST farms and the Integrated Disposal Facility (IDF).

  17. Remaining Sites Verification Package for the 331 Life Sciences Laboratory Drain Field Septic System, Waste Site Reclassification Form 2008-020

    SciTech Connect

    J. M. Capron

    2008-10-16

    The 331 Life Sciences Laboratory Drain Field (LSLDF) septic system waste site consists of a diversion chamber, two septic tanks, a distribution box, and a drain field. This septic system was designed to receive sanitary waste water, from animal studies conducted in the 331-A and 331-B Buildings, for discharge into the soil column. However, field observations and testing suggest the 331 LSLDF septic system did not receive any discharges. In accordance with this evaluation, the confirmatory sampling results support a reclassification of the 331 LSLDF waste site to No Action. This site does not have a deep zone or other condition that would warrant an institutional control in accordance with the 300-FF-2 ROD under the industrial land use scenario.

  18. Remaining Sites Verification Package for the 116-F-16, PNL Outfall and the 100-F-43, PNL Outfall Spillway, Waste Site Reclassification Form 2006-046

    SciTech Connect

    L. M. Dittmer

    2006-09-14

    The 100-F-43 waste site is the portion of the former discharge spillway for the PNL Outfall formerly existing above the ordinary high water mark of the Columbia River. The spillway consisted of a concrete flume used to discharge waste effluents from the 100-F Experimental Animal Farm. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  19. Remaining Sites Verification Package for the 100-D-24, 119-D Sample Building Drywell, Waste Site Reclassification Form 2006-004

    SciTech Connect

    L. M. Dittmer

    2006-09-19

    The 100-D-24 Sample Building Drywell waste site was a drywell that received drainage from a floor drain in the 119-D Sample Building. Confirmatory sampling was conducted on November 3, 2005. The waste site meets the remedial action objectives specified in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  20. Optimal segmentation and packaging process

    DOEpatents

    Kostelnik, K.M.; Meservey, R.H.; Landon, M.D.

    1999-08-10

    A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D and D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded. 3 figs.