Sample records for waste package materials

  1. Waste Package Materials Performance Peer Review | Department...

    Broader source: Energy.gov (indexed) [DOE]

    Waste Package Materials Performance Peer Review Waste Package Materials Performance Peer Review A consensus peer review of the current technical basis and the planned experimental...

  2. Aqueous Corrosion Rates for Waste Package Materials

    SciTech Connect (OSTI)

    S. Arthur

    2004-10-08T23:59:59.000Z

    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.

  3. EVALUATION OF THE FINAL REPORT: WASTE PACKAGE MATERIALS PERFORMANCE...

    Broader source: Energy.gov (indexed) [DOE]

    EVALUATION OF THE FINAL REPORT: WASTE PACKAGE MATERIALS PERFORMANCE PEER REVIEW PANEL B00000000-01717-5700-00005 REV 00 August 2002 This document is not an official copy and is for...

  4. Waste disposal package

    DOE Patents [OSTI]

    Smith, M.J.

    1985-06-19T23:59:59.000Z

    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.

  5. Radioactive waste disposal package

    DOE Patents [OSTI]

    Lampe, Robert F. (Bethel Park, PA)

    1986-01-01T23:59:59.000Z

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

    SciTech Connect (OSTI)

    J.P. Nicot

    2000-09-29T23:59:59.000Z

    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 calculation supports disposal criticality analysis and has been prepared in accordance with AP-3.12Q, Calculations (Ref. 49). This calculation uses results from Ref. 4 on actinide accumulation in the invert and more generally does reference heavily the cited calculation. In addition to the information provided in this calculation, the reader is referred to the cited calculation for a more thorough treatment of items applying to both the invert and fracture system such as the choice of the thermodynamic database, the composition of J-13 well water, tuff composition, dissolution rate laws, Pu(OH){sub 4} solubility and also for details on the source term composition. The flow conditions (seepage rate, water velocity in fractures) in the drift and the fracture system beneath initially referred to the TSPA-VA because this work was prepared before the release of the work feeding the TSPA-SR. Some new information feeding the TSPA-SR has since been included. Similarly, the soon-to-be-qualified thermodynamic database data0.ymp has not been released yet.

  7. The role of laboratory analog experiments in assessing the performance of waste package materials

    SciTech Connect (OSTI)

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

    1990-12-31T23:59:59.000Z

    There is an immediate need to begin to validate models that can be used for assessing the performance of waste package materials in an unsaturated repository environment. This paper examines available testing information and testing approaches that could support validation of models for engineering barrier system (EBS) radionuclide release. The content is presented in the context of the general methodology that has been proposed for validating performance assessment models. Available experimental observations are used to test some of the EBS release rate modeling premises. These observations include evidence of fluid film formation on waste glass surfaces in isothermal humid environments, accelerated waste glass reaction rates under repository service conditions of large glass surface area to water volume ratio, and mobilization of radionuclides as solutes and colloids. It is concluded that some important modeling premises may not be consistent with available experimental information. However, it is also concluded that future laboratory testing, which simulates the integrated waste package systems, is needed to evaluate the significance of these inconsistencies and to test the system level models. A small-scale apparatus which was developed and tested to examine the feasibility of laboratory analog testing for the unsaturated Yucca Mountain repository environment is described. 16 refs., 4 figs., 1 tab.

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

    SciTech Connect (OSTI)

    Romano, T.

    1997-09-25T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    BOEHNKE, W.M.

    2000-07-13T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Adrados, A., E-mail: aitziber.adrados@ehu.es [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain); De Marco, I.; Lopez-Urionabarrenechea, A.; Caballero, B.M.; Laresgoiti, M.F. [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain)

    2013-01-15T23:59:59.000Z

    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.

  11. Tritium waste package

    DOE Patents [OSTI]

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

    1995-11-07T23:59:59.000Z

    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.

  12. Tritium waste package

    DOE Patents [OSTI]

    Rossmassler, Rich (Cranbury, NJ); Ciebiera, Lloyd (Titusville, NJ); Tulipano, Francis J. (Teaneck, NJ); Vinson, Sylvester (Ewing, NJ); Walters, R. Thomas (Lawrenceville, NJ)

    1995-01-01T23:59:59.000Z

    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.

  13. Nuclear Material Packaging Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2008-03-07T23:59:59.000Z

    The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. No cancellation. Certified 11-18-10.

  14. Radioactive material package seal tests

    SciTech Connect (OSTI)

    Madsen, M.M.; Humphreys, D.L.; Edwards, K.R.

    1990-01-01T23:59:59.000Z

    General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the US Code of Federal Regulations Part 71 (US Nuclear Regulatory Commission, 1983). The requirements for Type B packages provide a broad range of environments under which the system must contain the RAM without posing a threat to health or property. Seals that provide the containment system interface between the packaging body and the closure must function in both high- and low-temperature environments under dynamic and static conditions. A seal technology program, jointly funded by the US Department of Energy Office of Environmental Restoration and Waste Management (EM) and the Office of Civilian Radioactive Waste Management (OCRWM), was initiated at Sandia National Laboratories. Experiments were performed in this program to characterize the behavior of several static seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Because most elastomer O-ring applications are for hydraulic systems, manufacturer low-temperature ratings are based on methods that simulate this use. The seal materials tested in this program with a fixture similar to a RAM cask closure, with the exception of silicone S613-60, are not leak tight (1.0 {times} 10{sup {minus}7} std cm{sup 3}/s) at manufacturer low-temperature ratings. 8 refs., 3 figs., 1 tab.

  15. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    SciTech Connect (OSTI)

    S. LeStrange

    1999-11-15T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1998-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2006-04-01T23:59:59.000Z

    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.

  19. Packaging - Materials review

    SciTech Connect (OSTI)

    Herrmann, Matthias [Hoppecke Advanced Battery Technology GmbH, 08056 Zwickau (Germany)

    2014-06-16T23:59:59.000Z

    Nowadays, a large number of different electrochemical energy storage systems are known. In the last two decades the development was strongly driven by a continuously growing market of portable electronic devices (e.g. cellular phones, lap top computers, camcorders, cameras, tools). Current intensive efforts are under way to develop systems for automotive industry within the framework of electrically propelled mobility (e.g. hybrid electric vehicles, plug-in hybrid electric vehicles, full electric vehicles) and also for the energy storage market (e.g. electrical grid stability, renewable energies). Besides the different systems (cell chemistries), electrochemical cells and batteries were developed and are offered in many shapes, sizes and designs, in order to meet performance and design requirements of the widespread applications. Proper packaging is thereby one important technological step for designing optimum, reliable and safe batteries for operation. In this contribution, current packaging approaches of cells and batteries together with the corresponding materials are discussed. The focus is laid on rechargeable systems for industrial applications (i.e. alkaline systems, lithium-ion, lead-acid). In principle, four different cell types (shapes) can be identified - button, cylindrical, prismatic and pouch. Cell size can be either in accordance with international (e.g. International Electrotechnical Commission, IEC) or other standards or can meet application-specific dimensions. Since cell housing or container, terminals and, if necessary, safety installations as inactive (non-reactive) materials reduce energy density of the battery, the development of low-weight packages is a challenging task. In addition to that, other requirements have to be fulfilled: mechanical stability and durability, sealing (e.g. high permeation barrier against humidity for lithium-ion technology), high packing efficiency, possible installation of safety devices (current interrupt device, valve, etc.), chemical inertness, cost issues, and others. Finally, proper cell design has to be considered for effective thermal management (i.e. cooling and heating) of battery packs.

  20. 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 (OSTI)

    Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (USA); Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S. [CDA/INCRA Joint Advisory Group, Greenwich, CT (USA)

    1990-06-01T23:59:59.000Z

    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.

  1. CRAD, Packaging and Transfer of Hazardous Materials and Materials...

    Office of Environmental Management (EM)

    CRAD, Packaging and Transfer of Hazardous Materials and Materials of National Security Interest Assessment Plan CRAD, Packaging and Transfer of Hazardous Materials and Materials of...

  2. Engineered waste-package-system design specification

    SciTech Connect (OSTI)

    Not Available

    1983-05-01T23:59:59.000Z

    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. CERAMIC WASTE FORM DATA PACKAGE

    SciTech Connect (OSTI)

    Amoroso, J.; Marra, J.

    2014-06-13T23:59:59.000Z

    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.

  4. Roadmapping the Resolution of Gas Generation Issues in Packages Containing Radioactive Waste/Materials - A Status Report

    SciTech Connect (OSTI)

    Luke, D.E. (INEEL); Hamp, S. (DOE-Albuquerque Operations Office)

    2002-01-04T23:59:59.000Z

    Gas generation issues, particularly hydrogen, have been an area of concern for the transport and storage of radioactive materials and waste in the Department of Energy (DOE) Complex. Potentially combustible gases can be generated through a variety of reactions, including chemical reactions and radiolytic decomposition of hydrogen-containing material. Since transportation regulations prohibit shipment of explosives and radioactive materials together, it was decided that hydrogen generation was a problem that warranted the execution of a high-level roadmapping effort. This paper discusses the major gas generation issues within the DOE Complex and the research that has been and is being conducted by the transuranic (TRU) waste, nuclear materials, and spent nuclear fuels (SNF) programs within DOE's Environmental Management (EM) organizations to address gas generation concerns. This paper presents a ''program level'' roadmap that links technology development to program needs and identifies the probability of success in an effort to understand the programmatic risk associated with the issue of gas generation. This paper also presents the status of the roadmap and follow-up activities.

  5. Roadmapping the Resolution of Gas Generation Issues in Packages Containing Radioactive Waste/Materials - A Status Report

    SciTech Connect (OSTI)

    Luke, Dale Elden; Hamp, S.

    2002-02-01T23:59:59.000Z

    Gas generation issues, particularly hydrogen, have been an area of concern for the transport and storage of radioactive materials and waste in the Department of Energy (DOE) Complex. Potentially combustible gases can be generated through a variety of reactions, including chemical reactions and radiolytic decomposition of hydrogen- containing material. Since transportation regulations prohibit shipment of explosives and radioactive materials together, it was decided that hydrogen generation was a problem that warranted the execution of a high-level roadmapping effort. This paper discusses the major gas generation issues within the DOE Complex and the research that has been and is being conducted by the transuranic (TRU) waste, nuclear materials, and spent nuclear fuels (SNF) programs within DOE’s Environmental Management (EM) organizations to address gas generation concerns. This paper presents a "program level" roadmap that links technology development to program needs and identifies the probability of success in an effort to understand the programmatic risk associated with the issue of gas generation. This paper also presents the status of the roadmap and follow-up activities.

  6. Roadmapping - A Tool for Resolving Science and Technology Issues Related to Processing, Packaging, and Shipping Nuclear Materials and Waste

    SciTech Connect (OSTI)

    Luke, Dale Elden; Dixon, Brent Wayne; Murphy, James Anthony

    2002-06-01T23:59:59.000Z

    Roadmapping is an effective methodology to identify and link technology development and deployment efforts to a program's or project's needs and requirements. Roadmapping focuses on needed technical support to the baselines (and to alternatives to the baselines) where the probability of success is low (high uncertainty) and the consequences of failure are relatively high (high programmatic risk, higher cost, longer schedule, or higher ES&H risk). The roadmap identifies where emphasis is needed, i.e., areas where investments are large, the return on investment is high, or the timing is crucial. The development of a roadmap typically involves problem definition (current state versus the desired state) and major steps (functions) needed to reach the desired state. For Nuclear Materials (NM), the functions could include processing, packaging, storage, shipping, and/or final disposition of the material. Each function is examined to determine what technical development would be needed to make the function perform as desired. This requires a good understanding of the current state of technology and technology development and validation activities to ensure the viability of each step. In NM disposition projects, timing is crucial! Technology must be deployed within the project window to be of value. Roadmaps set the stage to keep the technology development and deployment focused on project milestones and ensure that the technologies are sufficiently mature when needed to mitigate project risk and meet project commitments. A recent roadmapping activity involved a 'cross-program' effort, which included NM programs, to address an area of significant concern to the Department of Energy (DOE) related to gas generation issues, particularly hydrogen. The roadmap that was developed defined major gas generation issues within the DOE complex and research that has been and is being conducted to address gas generation concerns. The roadmap also provided the basis for sharing ''lessons learned'' from R&D efforts across DOE programs to increase efficiency and effectiveness in addressing gas generation issues. The gas generation roadmap identified pathways that have significant risk, indicating where more emphasis should be placed on contingency planning. Roadmapping further identified many opportunities for sharing of information and collaboration. Roadmapping will continue to be useful in keeping focused on the efforts necessary to mitigate the risk in the disposition pathways and to respond to the specific needs of the sites. Other areas within NM programs, including transportation and disposition of orphan and other nuclear materials, are prime candidates for additional roadmapping to assure achievement of timely and cost effective solutions for the processing, packaging, shipping, and/or final disposition of nuclear materials.

  7. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-11-18T23:59:59.000Z

    Establishes requirements and responsibilities for management of Department of Energy (DOE), including National Nuclear Security Administration, materials transportation and packaging and ensures the safe, secure, efficient packaging and transportation of materials, both hazardous and non-hazardous.

  8. Strategy for experimental validation of waste package performance assessment

    SciTech Connect (OSTI)

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

    1990-07-01T23:59:59.000Z

    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.

  9. Hanford Site radioactive hazardous materials packaging directory

    SciTech Connect (OSTI)

    McCarthy, T.L.

    1995-12-01T23:59:59.000Z

    The Hanford Site Radioactive Hazardous Materials Packaging Directory (RHMPD) provides information concerning packagings owned or routinely leased by Westinghouse Hanford Company (WHC) for offsite shipments or onsite transfers of hazardous materials. Specific information is provided for selected packagings including the following: general description; approval documents/specifications (Certificates of Compliance and Safety Analysis Reports for Packaging); technical information (drawing numbers and dimensions); approved contents; areas of operation; and general information. Packaging Operations & Development (PO&D) maintains the RHMPD and may be contacted for additional information or assistance in obtaining referenced documentation or assistance concerning packaging selection, availability, and usage.

  10. ALTERNATE MATERIALS IN DESIGN OF RADIOACTIVE MATERIAL PACKAGES

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2010-07-09T23:59:59.000Z

    This paper presents a summary of design and testing of material and composites for use in radioactive material packages. These materials provide thermal protection and provide structural integrity and energy absorption to the package during normal and hypothetical accident condition events as required by Title 10 Part 71 of the Code of Federal Regulations. Testing of packages comprising these materials is summarized.

  11. Drift emplaced waste package thermal response

    SciTech Connect (OSTI)

    Ruffner, D.J.; Johnson, G.L.; Platt, E.A.; Blink, J.A. [Lawrence Livermore National Lab., CA (United States); Doering, T.W. [B and W Fuel Co., Lynchburg, VA (United States)

    1993-01-01T23:59:59.000Z

    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.

  12. Hazardous Material Packaging for Transport - Administrative Procedures

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1986-09-30T23:59:59.000Z

    To establ1sh administrative procedures for the certification and use of radioactive and other hazardous materials packaging by the Department of Energy (DOE).

  13. Secondary Waste Form Down Selection Data Package – Ceramicrete

    SciTech Connect (OSTI)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31T23:59:59.000Z

    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 binder is formed through an acid-base reaction between calcined magnesium oxide (MgO; a base) and potassium hydrogen phosphate (KH{sub 2}PO{sub 4}; an acid) in aqueous solution. The reaction product sets at room temperature to form a highly crystalline material. During the reaction, the hazardous and radioactive contaminants also react with KH{sub 2}PO{sub 4} to form highly insoluble phosphates. In this data package, physical property and waste acceptance data for Ceramicrete waste forms fabricated with wastes having compositions that were similar to those expected for secondary waste effluents, as well as secondary waste effluent simulants from the Hanford Tank Waste Treatment and Immobilization Plant were reviewed. With the exception of one secondary waste form formulation (25FA+25 W+1B.A. fabricated with the mixed simulant did not meet the compressive strength requirement), all the Ceramicrete waste forms that were reviewed met or exceeded Integrated Disposal Facility waste acceptance criteria.

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

    SciTech Connect (OSTI)

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

    1982-02-01T23:59:59.000Z

    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.

  15. Packaging and Transfer of Hazardous Materials and Materials of...

    Broader source: Energy.gov (indexed) [DOE]

    PACKAGING AND TRANSFER OF HAZARDOUS MATERIALS AND MATERIALS OF NATIONAL SECURITY INTEREST Assessment Plan NNSANevada Site Office Facility Representative Division Performance...

  16. Industrial Waste Landfill IV upgrade package

    SciTech Connect (OSTI)

    Not Available

    1994-03-29T23:59:59.000Z

    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.

  17. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2004-12-22T23:59:59.000Z

    The Order establishes requirements and responsibilities for management of Department of Energy (DOE), including National Nuclear Security Administration (NNSA), materials transportation and packaging to ensure the safe, secure, efficient packaging and transportation of materials, both hazardous and nonhazardous. Cancels DOE O 460.2 and DOE O 460.2 Chg 1

  18. RECLAMATION OF RADIOACTIVE MATERIAL PACKAGING COMPONENTS

    SciTech Connect (OSTI)

    Abramczyk, G.; Nathan, S.; Loftin, B.; Bellamy, S.

    2011-06-06T23:59:59.000Z

    Radioactive material packages are withdrawn from use for various reasons; loss of mission, decertification, damage, replacement, etc. While the packages themselves may be decertified, various components may still be able to perform to their required standards and find useful service. The Packaging Technology and Pressurized Systems group of the Savannah River National Laboratory has been reducing the cost of producing new Type B Packagings by reclaiming, refurbishing, and returning to service the containment vessels from older decertified packagings. The program and its benefits are presented.

  19. Carbon Foam Thermal Management Materials for Electronic Packaging...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Foam Thermal Management Materials for Electronic Packaging Carbon Foam Thermal Management Materials for Electronic Packaging Presentation from the U.S. DOE Office of Vehicle...

  20. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-09-27T23:59:59.000Z

    Establishes Department of Energy (DOE) policies and requirements to supplement applicable laws, rules, regulations, and other DOE Orders for materials transportation and packaging operations. Cancels DOE 1540.1A, DOE 1540.2, DOE 1540.3A.

  1. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-10-26T23:59:59.000Z

    Establishes Department of Energy (DOE) policies and requirements to supplement applicable laws, rules, regulations, and other DOE Orders for materials transportation and packaging operations. Cancels: DOE 1540.1A, DOE 1540.2, and DOE 1540.3A.

  2. Hazardous Materials Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2015-04-20T23:59:59.000Z

    The Order establishes safety requirements for the proper packaging and transportation of Department of offsite shipments and onsite transfers of radioactive and other hazardous materials, and for modal transportation.

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

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2013-10-10T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    D.S. Kimball; C.E. Sanders

    2006-02-07T23:59:59.000Z

    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.

  5. The radioactive materials packaging handbook: Design, operations, and maintenance

    SciTech Connect (OSTI)

    Shappert, L.B.; Bowman, S.M. [Oak Ridge National Lab., TN (United States); Arnold, E.D. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States)] [and others

    1998-08-01T23:59:59.000Z

    As part of its required activities in 1994, the US Department of Energy (DOE) made over 500,000 shipments. Of these shipments, approximately 4% were hazardous, and of these, slightly over 1% (over 6,400 shipments) were radioactive. Because of DOE`s cleanup activities, the total quantities and percentages of radioactive material (RAM) that must be moved from one site to another is expected to increase in the coming years, and these materials are likely to be different than those shipped in the past. Irradiated fuel will certainly be part of the mix as will RAM samples and waste. However, in many cases these materials will be of different shape and size and require a transport packaging having different shielding, thermal, and criticality avoidance characteristics than are currently available. This Handbook provides guidance on the design, testing, certification, and operation of packages for these materials.

  6. Contact-Handled and Remote-Handled Transuranic Waste Packaging

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2011-08-09T23:59:59.000Z

    Provides specific instructions for packaging and/or repackaging contact-handled transuranic (CH-TRU) and remote-handled transuranic (RH-TRU) waste in a manner consistent with DOE O 435.1, Radioactive Waste Management, DOE M 435.1-1 Chg 1, Radioactive Waste Management Manual, CH-TRU and RH-TRU waste transportation requirements, and Waste Isolation Pilot Plant (WIPP) programmatic requirements. Does not cancel other directives.

  7. EM Waste and Materials Disposition & Transportation | Department...

    Office of Environmental Management (EM)

    EM Waste and Materials Disposition & Transportation EM Waste and Materials Disposition & Transportation DOE's Radioactive Waste Management Priorities: Continue to manage waste...

  8. Ceramic package fabrication for YMP nuclear waste disposal

    SciTech Connect (OSTI)

    Wilfinger, K.

    1994-08-01T23:59:59.000Z

    The purpose of this work is to develop alternate materials/design concepts to metal barriers for the Nevada Nuclear Waste Storage Investigations Project. There is some potential that site conditions may prove to be too aggressive for successful employment of the metal alloys under current consideration or that performance assessment models will predict metal container degradation rates that are inconsistent with the goal of substantially complete containment included in the NRC regulations. In the event that the anticipated lifetimes of metal containers are considered inadequate, alternate materials (i.e. ceramics or ceramic/metal composites) will be chosen due to superior corrosion resistance. This document was prepared using information taken from the open literature, conversations and correspondence with vendors, news releases and data presented at conferences to determine what form such a package might take. This discussion presents some ceramic material selection criteria, alternatives for the materials which might be used and alternatives for potential fabrication routes. This includes {open_quotes}stand alone{close_quotes} ceramic components and ceramic coatings/linings for metallic structures. A list of companies providing verbal or written information concerning the production of ceramic or ceramic lined waste containers appears at the end of this discussion.

  9. RADIOACTIVE MATERIAL PACKAGING TORQUE REQUIREMENTS COMPLIANCE

    SciTech Connect (OSTI)

    Watkins, R.; Leduc, D.

    2011-03-24T23:59:59.000Z

    Shipping containers used to transport radioactive material (RAM) in commerce employ a variety of closure mechanisms. Often, these closure mechanisms require a specific amount of torque be applied to a bolt, nut or other threaded fastener. It is important that the required preload is achieved so that the package testing and analysis is not invalidated for the purpose of protecting the public. Torque compliance is a means of ensuring closure preload, is a major factor in accomplishing the package functions of confinement/containment, sub-criticality, and shielding. This paper will address the importance of applying proper torque to package closures, discuss torque value nomenclature, and present one methodology to ensure torque compliance is achieved.

  10. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-05-31T23:59:59.000Z

    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).

  11. Motor Packaging with Consideration of Electromagnetic and Material...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Documents & Publications Motor Packaging with Consideration of Electromagnetic and Material Characteristics Alnico and Ferrite Hybrid Excitation Electric Machines Wireless Charging...

  12. Review of waste package verification tests. Semiannual report, October 1983-March 1984. Volume 4

    SciTech Connect (OSTI)

    Jain, H.; Veakis, E.; Soo, P.

    1985-06-01T23:59:59.000Z

    The current study is part of an ongoing task to specify tests that may be used to verify that engineered waste package/repository systems comply with NRC radionuclide containment and controlled release performance objectives. Work covered in this report includes crushed tuff packing material for use in a high-level waste tuff repository. Ranges of repository conditions relevant to its testing and other factors important for its performance are discussed. 23 refs., 5 figs., 3 tabs.

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

    SciTech Connect (OSTI)

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

    1987-10-01T23:59:59.000Z

    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.

  14. Quality assurance for radioactive waste packages -- A general approach

    SciTech Connect (OSTI)

    Martens, B.R. [Bundesamt fuer Strahlenschutz, Saltzgitter (Germany)

    1993-12-31T23:59:59.000Z

    Radioactive waste packages must fulfill the requirements resulting from regulations concerning handling, treatment, conditioning, transportation, storage and disposal so that the goal of radioactive waste management can be achieved. Usually in different parts of waste management different quality systems are used, and different quality assurance measures are performed. In the paper, these problems ar elucidated and it is explained by means of the quality assurance performed for the disposal of radioactive waste in Germany how the fulfillment of the requirements of the repository can be ensured.

  15. WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT

    SciTech Connect (OSTI)

    M. C. Knapp

    1999-09-23T23:59:59.000Z

    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.

  16. Measurement of radionuclides in waste packages

    DOE Patents [OSTI]

    Brodzinski, Ronald L. (Richland, WA); Perkins, Richard W. (Richland, WA); Rieck, Henry G. (Richland, WA); Wogman, Ned A. (Richland, WA)

    1986-01-01T23:59:59.000Z

    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.

  17. Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

    SciTech Connect (OSTI)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1989-09-01T23:59:59.000Z

    This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100{degree}C. 52 refs., 9 figs.

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

    SciTech Connect (OSTI)

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

    1995-12-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2001-05-23T23:59:59.000Z

    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.

  20. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect (OSTI)

    Forsberg, Charles W.; Dole, Leslie R. [Nuclear Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN, 37831-6165 (United States)

    2007-07-01T23:59:59.000Z

    Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can out-perform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal conductivity cask. A multi-year, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports. (authors)

  1. Radioactive waste material melter apparatus

    DOE Patents [OSTI]

    Newman, Darrell F. (Richland, WA); Ross, Wayne A. (Richland, WA)

    1990-01-01T23:59:59.000Z

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another.

  2. Radioactive waste material melter apparatus

    DOE Patents [OSTI]

    Newman, D.F.; Ross, W.A.

    1990-04-24T23:59:59.000Z

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

  3. Aging and Phase Stability of Waste Package Outer Barrier

    SciTech Connect (OSTI)

    F. Wong

    2004-09-28T23:59:59.000Z

    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 present in the weld metal in the as-welded condition. It may be possible to eliminate these phases through a solution anneal heat treatment, but that may not be possible for the closure weld because the spent nuclear fuel cladding cannot be heated to more than 350 C. The effects of any stress mitigation techniques (such as laser peening or solution heat treating) that may be used to reduce the tensile stresses on the closure welds must also be determined. Cold-work will cause an increase in dislocation density, and such an increase in dislocation density may cause an increase in diffusion rates that control precipitation kinetics (Porter et al. 1992 [DIRS 161265]; Tawancy et al. 1983 [DIRS 104991]). Long-range order (LRO) occurs in nickel (Ni)-Cr-Mo alloys (such as Alloy 22) at temperatures less than approximately 600 C. This ordering has been linked to an increased susceptibility of Ni-Cr-Mo alloys to stress corrosion cracking and hydrogen embrittlement (Tawancy et al. 1983 [DIRS 104991]). These analyses provide information on the rate at which LRO may occur in Alloy 22 under repository conditions. Determination of the kinetics of transformations through experimental techniques requires that the transformations being investigated be accelerated due to the fact that the expected service life is at least 10,000 years. Phase transformations are typically accelerated through an increase in temperature. The rate of transformation is determined at the higher temperature and is extrapolated to the lower temperatures of interest.

  4. Advanced Ceramic Materials and Packaging Technologies for Realizing...

    Broader source: Energy.gov (indexed) [DOE]

    Thermal Hydraulic Laboratory Project start date: November 15, 2012 Advanced Ceramic Materials and Packaging Technologies for Realizing Sensors for Concentrating Solar Power...

  5. Base Technology for Radioactive Material Transportation Packaging Systems

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1992-07-08T23:59:59.000Z

    To establish Department of Energy (DOE) policies and responsibilities for coordinating and planning base technology for radioactive material transportation packaging systems.

  6. Carbon Foam Thermal Management Materials for Electronic Packaging...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Carbon Foam Thermal Management Materials for Electronic Packaging David P. Stinton Oak Ridge National Laboratory Project Team: Nidia C. Gallego, ORNL Brian Thompson, ThermalCentric...

  7. Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

    SciTech Connect (OSTI)

    Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

    2011-09-12T23:59:59.000Z

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  8. Number of Waste Package Hit by Igneous Intrusion

    SciTech Connect (OSTI)

    M. Wallace

    2004-10-13T23:59:59.000Z

    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.

  9. Radioactive waste material disposal

    DOE Patents [OSTI]

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1995-10-24T23:59:59.000Z

    The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide. 3 figs.

  10. Radioactive waste material disposal

    DOE Patents [OSTI]

    Forsberg, Charles W. (155 Newport Dr., Oak Ridge, TN 37830); Beahm, Edward C. (106 Cooper Cir., Oak Ridge, TN 37830); Parker, George W. (321 Dominion Cir., Knoxville, TN 37922)

    1995-01-01T23:59:59.000Z

    The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide.

  11. Electrochemical corrosion studies on copper-base waste package container materials in unirradiated 0.1 N NaNO{sub 3} at 95{degrees}C

    SciTech Connect (OSTI)

    Akkaya, M.; Verink, E.D. Jr. [Florida Univ., Gainesville, FL (USA); Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (USA)

    1988-05-01T23:59:59.000Z

    Three candidate materials were investigated in this study in terms of their electrochemical corrosion behavior in unirradiated 0.1 N NaNO{sub 3} solutions at 95{degrees}C. Anodic polarization experiments were conducted to determine the passive current densities, pitting potentials, and other parameters, together with Cyclic Current Reversal Voltammetry tests to evaluate the stability and protectiveness of the passive oxides formed. X-ray diffraction and Auger Electron Spectroscopy were used for identification of the corrosion products as well as Scanning Electron Microscopy for the surface morphology studies. 2 refs., 22 figs., 2 tabs.

  12. Packaging material for thin film lithium batteries

    DOE Patents [OSTI]

    Bates, John B. (116 Baltimore Dr., Oak Ridge, TN 37830); Dudney, Nancy J. (11634 S. Monticello Rd., Knoxville, TN 37922); Weatherspoon, Kim A. (223 Wadsworth Pl., Oak Ridge, TN 37830)

    1996-01-01T23:59:59.000Z

    A thin film battery including components which are capable of reacting upon exposure to air and water vapor incorporates a packaging system which provides a barrier against the penetration of air and water vapor. The packaging system includes a protective sheath overlying and coating the battery components and can be comprised of an overlayer including metal, ceramic, a ceramic-metal combination, a parylene-metal combination, a parylene-ceramic combination or a parylene-metal-ceramic combination.

  13. Assessment of Quality Assurance Measures for Radioactive Material Transport Packages not Requiring Competent Authority Design Approval - 13282

    SciTech Connect (OSTI)

    Komann, Steffen; Groeke, Carsten; Droste, Bernhard [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44-46, 12203 Berlin (Germany)] [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44-46, 12203 Berlin (Germany)

    2013-07-01T23:59:59.000Z

    The majority of transports of radioactive materials are carried out in packages which don't need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP's leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the 'non-competent authority approved package types' are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment by the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, Chap. 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the 'regulatory density' for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)

  14. Standard Guide for Packaging Materials for Foods to Be Irradiated

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2009-01-01T23:59:59.000Z

    1.1 This guide provides a format to assist producers and users of food packaging materials in selecting materials that have the desirable characteristics for their intended use and comply with applicable standards or government authorizations. It outlines parameters that should be considered when selecting food-contact packaging materials intended for use during irradiation of prepackaged foods and it examines the criteria for fitness for their use. 1.2 This guide identifies known regulations and regulatory frameworks worldwide pertaining to packaging materials for holding foods during irradiation; but it does not address all regulatory issues associated with the selection and use of packaging materials for foods to be irradiated. It is the responsibility of the user of this guide to determine the pertinent regulatory issues in each country where foods are to be irradiated and where irradiated foods are distributed. 1.3 This guide does not address all of the food safety issues associated with the synergisti...

  15. Waste Disposal Guide HOW TO PROPERLY DISPOSE OF WASTE MATERIALS

    E-Print Network [OSTI]

    Schaefer, Marcus

    Waste Disposal Guide HOW TO PROPERLY DISPOSE OF WASTE MATERIALS GENERATED AT DEPAUL UNIVERSITY.4 Hazardous Waste Defined p.5 Chemical Waste Procedure for Generating Departments p.6 o A of Containers p.8 o E. Disposal of Empty Containers p.8 o F. Storage of Waste Chemicals p.8,9 o G

  16. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    SciTech Connect (OSTI)

    J.S. Tang

    2001-05-03T23:59:59.000Z

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

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

    SciTech Connect (OSTI)

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

    1988-08-01T23:59:59.000Z

    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.

  18. Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20T23:59:59.000Z

    A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

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

    SciTech Connect (OSTI)

    M. J. Appel and J. M. Capron

    2007-07-25T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    J. M. Capron

    2008-01-21T23:59:59.000Z

    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.

  1. Packaging and Transfer of Materials of National Security Interest Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-09-29T23:59:59.000Z

    This Technical Manual establishes requirements for operational safety controls for onsite operations and provides Department of Energy (DOE) technical safety requirements and policy objectives for development of an Onsite Packaging and Transfer Program, pursuant to DOE O 461.1A, Packaging and Transfer or Transportation of Materials of National Security Interest. The DOE contractor must document this program in its Onsite Packaging and Transfer Manual/Procedures. Admin Chg 1, 7-26-05. Certified 2-2-07. Canceled by DOE O 461.2.

  2. Release rates from partitioning and transmutation waste packages

    SciTech Connect (OSTI)

    Lee, W.W.L. (Lawrence Berkeley Lab., CA (United States)); Choi, Jor-Shan (Lawrence Livermore National Lab., CA (United States))

    1991-12-01T23:59:59.000Z

    Partitioning the actinides in light-water reactor spent fuel and transmuting them in actinide-burning liquid-metal reactors has been proposed as a potential method for reducing the public risks from geologic disposal of nuclear waste. As a first step towards quantifying the benefits for waste disposal of actinide burning, we have calculated the release rates of key radionuclides from waste packages resulting from actinide burning, and compare them with release rates from LWR spent fuel destined for disposal at the potential repository at Yucca Mountain. The wet-drip water-contact mode has been used. Analytic methods and parameter values are very similar to those used for assessing Yucca Mountain as a potential repository. Once released, the transport characteristics of radionuclides will be largely determined by site geology. For the most important nuclides such as I-129 and {Tc}-99, which are undiminished by actinide-burning reactors, it is not surprising that actinide burning offers little reduction in releases. For important actinides such as Np-237 and Pu isotopes, which are reduced in inventory, the releases are not reduced because the release rates are proportional to solubility, rather than inventory.

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

    SciTech Connect (OSTI)

    MANN, F.M.

    2000-03-02T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2001-02-01T23:59:59.000Z

    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.

  5. Complex-wide representation of material packaged in 3013 containers

    SciTech Connect (OSTI)

    Narlesky, Joshua E.; Peppers, Larry G.; Friday, Gary P.

    2009-06-01T23:59:59.000Z

    The DOE sites packaging plutonium oxide materials packaged according to Department of Energy 3013 Standard (DOE-STD-3013) are responsible for ensuring that the materials are represented by one or more samples in the Materials Identification and Surveillance (MIS) program. The sites categorized most of the materials into process groups, and the remaining materials were characterized, based on the prompt gamma analysis results. The sites issued documents to identify the relationships between the materials packaged in 3013 containers and representative materials in the MIS program. These “Represented” documents were then reviewed and concurred with by the MIS Working Group. However, these documents were developed uniquely at each site and were issued before completion of sample characterization, small-scale experiments, and prompt gamma analysis, which provided more detailed information about the chemical impurities and the behavior of the material in storage. Therefore, based on the most recent data, relationships between the materials packaged in 3013 containers and representative materials in the MIS program been revised. With the prompt gamma analysis completed for Hanford, Rocky Flats, and Savannah River Site 3013 containers, MIS items have been assigned to the 3013 containers for which representation is based on the prompt gamma analysis results. With the revised relationships and the prompt gamma analysis results, a Master “Represented” table has been compiled to document the linkages between each 3013 container packaged to date and its representative MIS items. This table provides an important link between the Integrated Surveillance Program database, which contains information about each 3013 container to the MIS items database, which contains the characterization, prompt gamma data, and storage behavior data from shelf-life experiments for the representative MIS items.

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

    SciTech Connect (OSTI)

    Souza, Daiane Cristini B.; Gimenes Tessaro, Ana Paula; Vicente, Roberto [Nuclear and Energy Research Institute Brazil, Radioactive Waste Management Department IPEN/GRR, Sao Paulo. SP. (Brazil)] [Nuclear and Energy Research Institute Brazil, Radioactive Waste Management Department IPEN/GRR, Sao Paulo. SP. (Brazil)

    2013-07-01T23:59:59.000Z

    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)

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

    SciTech Connect (OSTI)

    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

    2003-09-11T23:59:59.000Z

    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.

  8. Analysis of waste package environment as related to initiation and propagation of localized corrosion

    SciTech Connect (OSTI)

    Walton, J.C.; Sridhar, N.; Cragnolino, G.; Nair, P. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

    1993-12-31T23:59:59.000Z

    Models are presented for prediction of waste package lifetime for the proposed high-level nuclear waste repository located at Yucca Mountain, Nevada. The waste package environment is expected to range from saturated, highly-alkaline brine to bicarbonate waters of low ionic strength. Localized corrosion will be predicted by comparing predicted corrosion potentials of the container with critical potentials for pitting and crevice corrosion. The critical potentials will be estimated by a combined program of experiments and mechanistic modeling.

  9. Waste Form Degradation Model Integration for Engineered Materials...

    Office of Environmental Management (EM)

    Waste Form Degradation Model Integration for Engineered Materials Performance Waste Form Degradation Model Integration for Engineered Materials Performance The collaborative...

  10. Material Recycling and Waste Disposal Document Control

    E-Print Network [OSTI]

    Guillas, Serge

    1 Material Recycling and Waste Disposal Procedure Document Control Document Created by 23, treatment, handling, transport and disposal of recyclable materials and residual wastes so as to maximise the opportunity and value for the recyclable materials and to minimise the quantity of residual materials

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

    SciTech Connect (OSTI)

    NA

    2005-10-25T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    SULLIVAN, T.

    2004-06-30T23:59:59.000Z

    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 distributed container failure times on cumulative release and peak concentration is provided to illustrate some of the complexities in safety assessment and how modeling can be used to support the conceptual approach in safety assessment and define data requirements. Two examples of the role of the waste form in controlling release are presented to illustrate the importance of waste form performance to safety assessment. These examples highlight the difficulties in changing the conceptual model from something that is conservative and defensible (such as instant release of all the activity) to more representative conceptual models that account for known physical and chemical processes (such as diffusion), The second waste form example accounts for the experimental observation that often a thin film with low diffusion properties forms on the waste form surface. The implications of formation of such a layer on release are investigated and the implications of attempting to account for this phenomena in a safety assessment are addressed.

  13. Treatment of halogen-containing waste and other waste materials

    DOE Patents [OSTI]

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1997-03-18T23:59:59.000Z

    A process is described for treating a halogen-containing waste material. The process provides a bath of molten glass containing a sacrificial metal oxide capable of reacting with a halogen in the waste material. The sacrificial metal oxide is present in the molten glass in at least a stoichiometric amount with respect to the halogen in the waste material. The waste material is introduced into the bath of molten glass to cause a reaction between the halogen in the waste material and the sacrificial metal oxide to yield a metal halide. The metal halide is a gas at the temperature of the molten glass. The gaseous metal halide is separated from the molten glass and contacted with an aqueous scrubber solution of an alkali metal hydroxide to yield a metal hydroxide or metal oxide-containing precipitate and a soluble alkali metal halide. The precipitate is then separated from the aqueous scrubber solution. The molten glass containing the treated waste material is removed from the bath as a waste glass. The process of the invention can be used to treat all types of waste material including radioactive wastes. The process is particularly suited for separating halogens from halogen-containing wastes. 3 figs.

  14. Treatment of halogen-containing waste and other waste materials

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN); Beahm, Edward C. (Oak Ridge, TN); Parker, George W. (Concord, TN)

    1997-01-01T23:59:59.000Z

    A process for treating a halogen-containing waste material. The process provides a bath of molten glass containing a sacrificial metal oxide capable of reacting with a halogen in the waste material. The sacrificial metal oxide is present in the molten glass in at least a stoichiometric amount with respect to the halogen in the waste material. The waste material is introduced into the bath of molten glass to cause a reaction between the halogen in the waste material and the sacrificial metal oxide to yield a metal halide. The metal halide is a gas at the temperature of the molten glass. The gaseous metal halide is separated from the molten glass and contacted with an aqueous scrubber solution of an alkali metal hydroxide to yield a metal hydroxide or metal oxide-containing precipitate and a soluble alkali metal halide. The precipitate is then separated from the aqueous scrubber solution. The molten glass containing the treated waste material is removed from the bath as a waste glass. The process of the invention can be used to treat all types of waste material including radioactive wastes. The process is particularly suited for separating halogens from halogen-containing wastes.

  15. Benchmarking of finite element codes for radioactive material transportation packages

    SciTech Connect (OSTI)

    Ammerman, D.J.

    1996-10-01T23:59:59.000Z

    The increased power of computers and computer codes makes the use of nonlinear dynamic finite element analyses attractive for use as a tool used in the design and certification of radioactive material transportation packages. For this analysis technique to be acceptable it must be demonstrated. The technique has the ability to accurately capture the response of the packages to accident environments required by the regulations. The best method of demonstrating this ability is via a series of benchmark analyses. In this paper three benchmark problems involving significant inelastic deformations will be discussed. One of the problems has been analyzed using many different finite element codes. The other two problems involve comparison of finite element calculations to the results form physical tests. The ability of the finite element method to accurately capture the response in these three problems indicates the method should be acceptable for radioactive material transportation package design and certification.

  16. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    SciTech Connect (OSTI)

    Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

    2002-02-26T23:59:59.000Z

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used.

  17. advanced packaging materials: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    packaging materials First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Eco Friendly Pharmaceutical...

  18. DIGITAL RADIOGRAPHY OF SPECIAL NUCLEAR MATERIAL TEST PACKAGES

    SciTech Connect (OSTI)

    HOWARD, BOYD

    2006-02-02T23:59:59.000Z

    The purpose of this document is to provide a brief introduction to digital radiography (DR), and a description of the DR configuration that was used to radiographically image the Special Nuclear Material (SNM) Test Packages before and after function tests that have been conducted. Also included are (1) Attachment 1, a comprehensive index that describes at which phase of the certification process that digital radiographic images were acquired, (2) digital radiographic images of each of the six packages at various stages of the certification process, and (3) Attachment 2, imaging instructions, that specify the setup procedures and detailed parameters of the DR imaging methodology that were used.

  19. Testing of Packaging Materials for Improved PV Module Reliability

    SciTech Connect (OSTI)

    Jorgensen, G. J.; Terwilliger, K. M.; Kempe, M. D.; McMahon, T. J.

    2005-02-01T23:59:59.000Z

    A number of candidate alternative encapsulant and soft backsheet materials have been evaluated in terms of their suitability for photovoltaic (PV) module packaging applications. Relevant properties, including interfacial adhesion and moisture transport, have been measured as a function of damp-heat (85 C/85% relative humidity) exposure. Based on these tests, promising new encapsulants with improved properties have been identified. Backsheets prepared by industry and at NREL have been found to provide varying levels of moisture ingress protection. To achieve significantly improved products, further development of these candidates is ongoing. The relative effectiveness of various packaging strategies to protect PV devices has also been investigated.

  20. Nondestructive assay and nondestructive examination of remote-handled transuranic waste at the ORNL waste handling and packaging plant

    SciTech Connect (OSTI)

    Schultz, F.J.; Caldwell, J.T. (Oak Ridge National Lab., TN (USA); Pajarito Scientific Corp. (USA))

    1989-01-01T23:59:59.000Z

    The purpose of this investigation is to examine the use of an electron linear accelerator (LINAC) in the performance of nondestructive assay (NDA) and nondestructive examination (NDE) measurements of remote-handled transuranic wastes. The system will be used to perform waste characterization and certification activities at the Oak Ridge National Laboratory's proposed Waste Handling and Packaging Plant. The NDA and NDE technologies which were developed for contact-handled wastes are inadequate to perform such measurements on high gamma and neutron dose-rate wastes. A single LINAC will provide the interrogating fluxes required for both NDA and NDE measurements of the wastes. 11 refs., 6 figs.

  1. In-situ vitrification of waste materials

    DOE Patents [OSTI]

    Powell, J.R.; Reich, M.; Barletta, R.

    1997-10-14T23:59:59.000Z

    A method for the in-situ vitrification of waste materials in a disposable can that includes an inner container and an outer container is disclosed. The method includes the steps of adding frit and waste materials to the inner container, removing any excess water, heating the inner container such that the frit and waste materials melt and vitrify after cooling, while maintaining the outer container at a significantly lower temperature than the inner container. The disposable can is then cooled to ambient temperatures and stored. A device for the in-situ vitrification of waste material in a disposable can is also disclosed. 7 figs.

  2. Greater-than-Class C low-level radioactive waste characterization. Appendix E-4: Packaging factors for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Quinn, G.; Grant, P.; Winberg, M.; Williams, K.

    1994-09-01T23:59:59.000Z

    This report estimates packaging factors for several waste types that are potential greater-than-Class C (GTCC) low-level radioactive waste (LLW). The packaging factor is defined as the volume of a GTCC LLW disposal container divided by the as-generated or ``unpackaged`` volume of the waste loaded into the disposal container. Packaging factors reflect any processes that reduce or increase an original unpackaged volume of GTCC LLW, the volume inside a waste container not occupied by the waste, and the volume of the waste container itself. Three values are developed that represent (a) the base case or most likely value for a packaging factor, (b) a high case packaging factor that corresponds to the largest anticipated disposal volume of waste, and (c) a low case packaging factor for the smallest volume expected. GTCC LLW is placed in three categories for evaluation in this report: activated metals, sealed sources, and all other waste.

  3. Operations to be Performed in the Waste Package Dry Remediation Cell

    SciTech Connect (OSTI)

    Norman E. Cole; Randy K. Elwood

    2003-10-01T23:59:59.000Z

    Describes planned and proposed operations for remediating damaged and/or out-of-compliance waste packages, casks, DPCs, overpacks, and containers at the Yucca Mountain Dry Transfer Facility.

  4. NEW APPROACH TO ADDRESSING GAS GENERATION IN RADIOACTIVE MATERIAL PACKAGING

    SciTech Connect (OSTI)

    Watkins, R; Leduc, D; Askew, N

    2009-06-25T23:59:59.000Z

    Safety Analysis Reports for Packaging (SARP) document why the transportation of radioactive material is safe in Type A(F) and Type B shipping containers. The content evaluation of certain actinide materials require that the gas generation characteristics be addressed. Most packages used to transport actinides impose extremely restrictive limits on moisture content and oxide stabilization to control or prevent flammable gas generation. These requirements prevent some users from using a shipping container even though the material to be shipped is fully compliant with the remaining content envelope including isotopic distribution. To avoid these restrictions, gas generation issues have to be addressed on a case by case basis rather than a one size fits all approach. In addition, SARP applicants and review groups may not have the knowledge and experience with actinide chemistry and other factors affecting gas generation, which facility experts in actinide material processing have obtained in the last sixty years. This paper will address a proposal to create a Gas Generation Evaluation Committee to evaluate gas generation issues associated with Safety Analysis Reports for Packaging material contents. The committee charter could include reviews of both SARP approved contents and new contents not previously evaluated in a SARP.

  5. Methane generation from waste materials

    DOE Patents [OSTI]

    Samani, Zohrab A. (Las Cruces, NM); Hanson, Adrian T. (Las Cruces, NM); Macias-Corral, Maritza (Las Cruces, NM)

    2010-03-23T23:59:59.000Z

    An organic solid waste digester for producing methane from solid waste, the digester comprising a reactor vessel for holding solid waste, a sprinkler system for distributing water, bacteria, and nutrients over and through the solid waste, and a drainage system for capturing leachate that is then recirculated through the sprinkler system.

  6. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages

    SciTech Connect (OSTI)

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

    2006-07-31T23:59:59.000Z

    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 salt in the castable refractory block and it is released over the same time period as the salt. Therefore, to limit the impact of precipitated Fe on the release of 99Tc, both the amount of precipitated Fe in the BV glass and the diameter of these particles should be minimized.

  7. 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 (OSTI)

    Not Available

    1983-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Interrante, C.G. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of High-Level Waste Management; Fraker, A.C.; Escalante, E. [National Inst. of Standards and Technology (MSEL), Gaithersburg, MD (United States). Metallurgy Div.

    1993-06-01T23:59:59.000Z

    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.

  9. RECERTIFICATION OF THE MODEL 9977 RADIOACTIVE MATERIAL PACKAGING

    SciTech Connect (OSTI)

    Abramczyk, G.; Bellamy, S.; Loftin, B.; Nathan, S.

    2013-06-05T23:59:59.000Z

    The Model 9977 Packaging was initially issued a Certificate of Compliance (CoC) by the Department of Energy’s Office of Environmental Management (DOE-EM) for the transportation of radioactive material (RAM) in the Fall of 2007. This first CoC was for a single radioactive material and two packing configurations. In the five years since that time, seven Addendums have been written to the Safety Analysis Report for Packaging (SARP) and five Letter Amendments have been written that have authorized either new RAM contents or packing configurations, or both. This paper will discuss the process of updating the 9977 SARP to include all the contents and configurations, including the addition of a new content, and its submittal for recertification.

  10. Recovery of fissile materials from nuclear wastes

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1999-01-01T23:59:59.000Z

    A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

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

    SciTech Connect (OSTI)

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

    1988-05-01T23:59:59.000Z

    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.

  12. Thermal testing of packages for transport of radioactive wastes

    SciTech Connect (OSTI)

    Koski, J.A.

    1994-12-31T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Kevin Skinner; Greg Housley; Colleen Shelton-Davis

    2011-11-01T23:59:59.000Z

    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.

  14. Greater-than-Class C low-level radioactive waste characterization. Appendix H: Packaging factors for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Quinn, G.; Grant, P.

    1991-08-01T23:59:59.000Z

    This report develops and presents estimates for a set of three values that represent a reasonable range for the packaging factors for several waste streams that are potential greater-than-Class C low-level radioactive waste. The packaging factor is defined as the volume of a greater-than-Class C low-level waste disposal container divided by the original, as-generated or ``unpackaged,`` volume of the wastes loaded into the disposal container. Packaging factors take into account any processes that reduce or increase an original unpackaged volume of a greater-than-Class C low-level radioactive waste, the volume inside a waste container not occupied by the waste, and the volume of the waste container itself. The three values developed represent (a) the base case or most likely value for a packaging factor, (b) a high case packaging factor that corresponds to the largest anticipated volume of waste for disposal, and (c) a low case packaging factor for the smallest volume expected. Three categories of greater-than-Class C low-level waste are evaluated in this report: activated metals, sealed sources, and all other wastes. Estimates of reasonable packaging factors for the low, base, and high cases for the specific waste streams in each category are shown in Table H-1.

  15. Tank 241-CX-70 waste removal and packaging

    SciTech Connect (OSTI)

    DuVon, D.K.

    1993-06-01T23:59:59.000Z

    Tank 241-CX-70, located on the Hanford Site in Washington State, is a 30,000 gal single-shell storage tank built in 1952 to hold high-level process waste from pilot tests of the reduction-oxidation process. In 1979 decommissioning operations were begun by pumping liquid waste from the tank to the double-shell tank (DST) 101-AY. Not all the waste was removed at that time. Approximately 10,300 gal of sludge remained. On September 25, 1987, operations were resumed to remove the remaining waste using a sluicing and pumping method. This report documents the final removal of waste from Tank 241-CX-70.

  16. Tank 241-CX-70 waste removal and packaging

    SciTech Connect (OSTI)

    DuVon, D.K.

    1993-01-01T23:59:59.000Z

    Tank 241-CX-70, located on the Hanford Site in Washington State, is a 30,000 gal single-shell storage tank built in 1952 to hold high-level process waste from pilot tests of the reduction-oxidation process. In 1979 decommissioning operations were begun by pumping liquid waste from the tank to the double-shell tank (DST) 101-AY. Not all the waste was removed at that time. Approximately 10,300 gal of sludge remained. On September 25, 1987, operations were resumed to remove the remaining waste using a sluicing and pumping method. This report documents the final removal of waste from Tank 241-CX-70.

  17. Module Design, Materials, and Packaging Research Team: Activities and Capabilities

    SciTech Connect (OSTI)

    McMahon, T. J.; del Cueto, J.; Glick, S.; Jorgensen, G.; Kempe, M.; Kennedy, C.; Pern, J.; Terwilliger, K

    2005-01-01T23:59:59.000Z

    Our team activities are directed at improving PV module reliability by incorporating new, more effective, and less expensive packaging materials and techniques. New and existing materials or designs are evaluated before and during accelerated environmental exposure for the following properties: (1) Adhesion and cohesion: peel strength and lap shear. (2) Electrical conductivity: surface, bulk, interface and transients. (3) Water vapor transmission: solubility and diffusivity. (4) Accelerated weathering: ultraviolet, temperature, and damp heat tests. (5) Module and cell failure diagnostics: infrared imaging, individual cell shunt characterization, coring. (6) Fabrication improvements: SiOxNy barrier coatings and enhanced wet adhesion. (7) Numerical modeling: Moisture ingress/egress, module and cell performance, and cell-to-frame leakage current. (8) Rheological properties of polymer encapsulant and sheeting materials. Specific examples will be described.

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

    SciTech Connect (OSTI)

    Ferreira da Cruz, Nuno, E-mail: nunocruz@ist.utl.pt; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

    2014-02-15T23:59:59.000Z

    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 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 Portugal the industry is paying local authorities more than just the incremental costs of recycling (full costs of selective collection and sorting minus the avoided costs). To provide a more definitive judgment on the fairness of the systems it will be necessary to assess the cost efficiency of waste management operators (and judge whether operators are claiming costs or eliciting “prices”)

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

    SciTech Connect (OSTI)

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

    1986-04-01T23:59:59.000Z

    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.

  20. Process Knowledge Summary Report for Materials and Fuels Complex Contact-Handled Transuranic Debris Waste

    SciTech Connect (OSTI)

    R. P. Grant; P. J. Crane; S. Butler; M. A. Henry

    2010-02-01T23:59:59.000Z

    This Process Knowledge Summary Report summarizes the information collected to satisfy the transportation and waste acceptance requirements for the transfer of transuranic (TRU) waste between the Materials and Fuels Complex (MFC) and the Advanced Mixed Waste Treatment Project (AMWTP). The information collected includes documentation that addresses the requirements for AMWTP and the applicable portion of their Resource Conservation and Recovery Act permits for receipt and treatment of TRU debris waste in AMWTP. This report has been prepared for contact-handled TRU debris waste generated by the Idaho National Laboratory at MFC. The TRU debris waste will be shipped to AMWTP for purposes of supercompaction. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU debris waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for waste originating from MFC.

  1. Secondary Waste Form Down-Selection Data Package—DuraLith

    SciTech Connect (OSTI)

    Mattigod, Shas V.; Westsik, Joseph H.

    2011-09-15T23:59:59.000Z

    This data package developed for the DuraLith wasteform includes information available in the open literature and from data obtained from testing currently underway. DuraLith is an alkali-activated geopolymer waste form developed by the Vitreous State Laboratory at The Catholic University of America (VSL-CUA) for encapsulating liquid radioactive waste. A DuraLith waste form developed for treating Hanford secondary waste liquids is prepared by alkali-activation of a mixture of ground blast furnace slag and metakaolinite with sand used as a filler material. Based on optimization tests, solid waste loading of {approx}7.5% and {approx}14.7 % has been achieved using the Hanford secondary waste S1 and S4 simulants, respectively. The Na loading in both cases is equivalent to {approx}6 M. Some of the critical parameters for the DuraLith process include, hydrogen generation and heat evolution during activator solution preparation using the waste simulant, heat evolution during and after mixing the activator solution with the dry ingredients, and a working window of {approx}20 minutes to complete the pouring of the DuraLith mixture into molds. Results of the most recent testing indicated that the working window can be extended to {approx}30 minutes if 75 wt% of the binder components, namely, blast furnace slag and metakaolin are replaced by Class F fly ash. A preliminary DuraLith process flow sheet developed by VSL-CUA for processing Hanford secondary waste indicated that 10 to 22 waste monoliths (each 48 ft3 in volume) can be produced per day. There are no current pilot-scale or full-scale DuraLith plants under construction or in operation; therefore, the cost of DuraLith production is unknown. The results of the non-regulatory leach tests, EPA Draft 1313 and 1316, Waste Simulant S1-optimized DuraLith specimens indicated that the concentrations of RCRA metals (Ag, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268.48. The data from the EPA draft 1315 leach test showed that LI values for COCs, namely 99Tc and I, ranged from 8.2 to 11.4 and 4.3 to 7.5, respectively. These values indicate that 99Tc meets the WAC LI requirement of 9.0 whereas, the LI values for I does not meet the WAC requirement of 11.0. Results of Toxicity Characteristic Leaching Procedure (TCLP)(EPA Method 1311) conducted on Waste Simulant S1-optimized DuraLith specimens, indicated that the concentrations of RCRA metals (Ag, As, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268.48. The data from the ANSI/ANS 16.1 leach test showed that LI values for COC, namely Re (as a Tc surrogate), ranged from 8.06 to 10.81. The LI value for another COC, namely I, was not measured in this test. The results of the compressive strength testing of Waste Simulant S1-optimized DuraLith specimens indicated that the monoliths were physically robust with compressive strengths ranging from 115.5 MPa (16757 psi) to 156.2 MPA (22667 psi).

  2. Packaging Materials and Design for Improved PV Module Reliability

    SciTech Connect (OSTI)

    Jorgensen, G.; Terwilliger, K.; Kempe, M.; Pern, J.; Glick, S.; del Cueto, J.; Kennedy, C.; McMahon, T.

    2005-01-01T23:59:59.000Z

    A number of candidate alternative encapsulant and soft backsheet materials have been evaluated in terms of their suitability for photovoltaic (PV) module packaging applications. Relevant properties, including peel strength as a function of damp heat exposure and permeability, have been measured. Based on these tests, promising new encapsulants with adhesion-promoting primers have been identified that result in improved properties. Test results for backsheets provided by industry and prepared at the National Renewable Energy Laboratory (NREL) have suggested strategies to achieve significantly improved products. The ability of glass/glass and glass/breathable backsheet constructions laminated with various encapsulant and/or edge seal materials to protect thin-film aluminum coatings deposited onto glass substrates was assessed. Glass/glass laminate constructions can trap harmful compounds that catalyze moisture-driven corrosion of the aluminum. Constructions with breathable backsheets allow higher rates of moisture ingress, but also allow egress of deleterious substances that can result in decreased corrosion.

  3. Plasma vitrification of waste materials

    DOE Patents [OSTI]

    McLaughlin, David F. (Oakmont, PA); Dighe, Shyam V. (North Huntingdon, PA); Gass, William R. (Plum Boro, PA)

    1997-01-01T23:59:59.000Z

    This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles.

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

    SciTech Connect (OSTI)

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

    2013-05-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1982-11-01T23:59:59.000Z

    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.

  6. Regulatory and extra-regulatory testing to demonstrate radioactive material packaging safety

    SciTech Connect (OSTI)

    Ammerman, D.J.

    1997-06-01T23:59:59.000Z

    Packages for the transportation of radioactive material must meet performance criteria to assure safety and environmental protection. The stringency of the performance criteria is based on the degree of hazard of the material being transported. Type B packages are used for transporting large quantities of radioisotopes (in terms of A{sub 2} quantities). These packages have the most stringent performance criteria. Material with less than an A{sub 2} quantity are transported in Type A packages. These packages have less stringent performance criteria. Transportation of LSA and SCO materials must be in {open_quotes}strong-tight{close_quotes} packages. The performance requirements for the latter packages are even less stringent. All of these package types provide a high level of safety for the material being transported. In this paper, regulatory tests that are used to demonstrate this safety will be described. The responses of various packages to these tests will be shown. In addition, the response of packages to extra-regulatory tests will be discussed. The results of these tests will be used to demonstrate the high level of safety provided to workers, the public, and the environment by packages used for the transportation of radioactive material.

  7. Nuclear materials stabilization and packaging. Quarterly progress report, October 1--December 31, 1995

    SciTech Connect (OSTI)

    Chidester, K.M. [comp.

    1996-05-01T23:59:59.000Z

    Progress is reported for Los Alamos Nuclear Materials Stabilization and Packaging projects for the first quarter of Fiscal Year 1996. Development and production activities in Plutonium Recovery and Processing, Plutonium Packaging, and Uranium Recovery and Processing are covered. Packaging quality assurance activities are reported.

  8. IN-PACKAGE CHEMISTRY ABSTRACTION

    SciTech Connect (OSTI)

    E. Thomas

    2005-07-14T23:59:59.000Z

    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.

  9. Buried waste containment system materials. Final Report

    SciTech Connect (OSTI)

    Weidner, J.R.; Shaw, P.G.

    1997-10-01T23:59:59.000Z

    This report describes the results of a test program to validate the application of a latex-modified cement formulation for use with the Buried Waste Containment System (BWCS) process during a proof of principle (POP) demonstration. The test program included three objectives. One objective was to validate the barrier material mix formulation to be used with the BWCS equipment. A basic mix formula for initial trials was supplied by the cement and latex vendors. The suitability of the material for BWCS application was verified by laboratory testing at the Idaho National Engineering and Environmental Laboratory (INEEL). A second objective was to determine if the POP BWCS material emplacement process adversely affected the barrier material properties. This objective was met by measuring and comparing properties of material prepared in the INEEL Materials Testing Laboratory (MTL) with identical properties of material produced by the BWCS field tests. These measurements included hydraulic conductivity to determine if the material met the US Environmental Protection Agency (EPA) requirements for barriers used for hazardous waste sites, petrographic analysis to allow an assessment of barrier material separation and segregation during emplacement, and a set of mechanical property tests typical of concrete characterization. The third objective was to measure the hydraulic properties of barrier material containing a stop-start joint to determine if such a feature would meet the EPA requirements for hazardous waste site barriers.

  10. Evaluation of the Compostability of Polymer-Coated Paperboard Packaging Materials

    E-Print Network [OSTI]

    Michel Jr., Frederick C.

    Evaluation of the Compostability of Polymer-Coated Paperboard Packaging Materials A report to the Heavy metals content of papers..........................................................5 Compost mix properties ......................................................................5 Composting

  11. Waste package performance assessment: Deterministic system model, program scope and specification

    SciTech Connect (OSTI)

    O`Connell, W.J.; Drach, R.S.

    1986-10-02T23:59:59.000Z

    Integrated assessments of the performance of nuclear waste package designs must be made in order to qualify waste package designs with respect to containment time and release-rate requirements. PANDORA is a computer-based model of the waste package and of the processes affecting it over the long terms, specific to conditions at the proposed Yucca Mountain, Nevada, site. The processes PANDORA models include: changes in inventories due to radioactive decay, gamma radiation dose rate in and near the package, heat transfer, mechanical behavior, groundwater contact, corrosion, waste form alteration, and radionuclide release. The model tracks the development and coupling of these processes over time. The process models are simplified ones that focus on major effects and on coupling. This report documents our conceptual model development and provides a specification for the computer program. The current model is the first in a series. Succeeding models will use guidance from results of preceding models in the PANDORA series and will incorporate results of recently completed experiments and calculations on processes affecting performance. 22 refs., 21 figs., 9 tabs.

  12. Economical Remediation of Plastic Waste into Advanced Materials...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Economical Remediation of Plastic Waste into Advanced Materials with Coatings Technology available for licensing: An autogenic pyrolysis process to convert plastic waste into...

  13. Nuclear Materials: Reconsidering Wastes and Assets - 13193

    SciTech Connect (OSTI)

    Michalske, T.A. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    The nuclear industry, both in the commercial and the government sectors, has generated large quantities of material that span the spectrum of usefulness, from highly valuable ('assets') to worthless ('wastes'). In many cases, the decision parameters are clear. Transuranic waste and high level waste, for example, have no value, and is either in a final disposition path today, or - in the case of high level waste - awaiting a policy decision about final disposition. Other materials, though discardable, have intrinsic scientific or market value that may be hidden by the complexity, hazard, or cost of recovery. An informed decision process should acknowledge the asset value, or lack of value, of the complete inventory of materials, and the structure necessary to implement the range of possible options. It is important that informed decisions are made about the asset value for the variety of nuclear materials available. For example, there is a significant quantity of spent fuel available for recycle (an estimated $4 billion value in the Savannah River Site's (SRS) L area alone); in fact, SRS has already blended down more than 300 metric tons of uranium for commercial reactor use. Over 34 metric tons of surplus plutonium is also on a path to be used as commercial fuel. There are other radiological materials that are routinely handled at the site in large quantities that should be viewed as strategically important and / or commercially viable. In some cases, these materials are irreplaceable domestically, and failure to consider their recovery could jeopardize our technological leadership or national defense. The inventories of nuclear materials at SRS that have been characterized as 'waste' include isotopes of plutonium, uranium, americium, and helium. Although planning has been performed to establish the technical and regulatory bases for their discard and disposal, recovery of these materials is both economically attractive and in the national interest. (authors)

  14. Method for recovering materials from waste

    DOE Patents [OSTI]

    Wicks, G.G.; Clark, D.E.; Schulz, R.L.

    1994-01-01T23:59:59.000Z

    A method for recovering metals from metals-containing wastes, a vitrifying the remainder of the wastes for disposal. Metals-containing wastes such as circuit boards, cathode ray tubes, vacuum tubes, transistors and so forth, are broken up and placed in a suitable container. The container is heated by microwaves to a first temperature in the range of approximately 300--800{degrees}C to combust organic materials in the waste, then heated further to a second temperature in the range of approximately 1000--1550{degrees}C at which temperature glass formers present in the waste will cause it to melt and vitrify. Low-melting-point metals such as tin and aluminum can be recovered after organics combustion is substantially complete. Metals with higher melting points, such as gold, silver and copper, can be recovered from the solidified product or separated from the waste at their respective melting points. Network former-containing materials can be added at the start of the process to assist vitrification.

  15. THERMAL EVALUATION OF DRUM TYPE RADIOACTIVE MATERIAL PACKAGING ARRAYS IN STORAGE

    SciTech Connect (OSTI)

    Gupta, N

    2009-04-27T23:59:59.000Z

    Drum type packages are routinely used to transport radioactive material (RAM) in the U.S. Department of Energy (DOE) complex. These packages are designed to meet the federal regulations described in 10 CFR 71.[1] In recent years, there has been a greater need to use these packagings to store the excess fissile material, especially plutonium for long term storage. While the design requirements for safe transportation of these packagings are well defined, the requirements for safe long term storage are not well established. Since the RAM contents in the packagings produce decay heat, it is important that they are stored carefully to prevent overheating of the containment vessel (CV) seals to prevent any leakage and the impact limiter to maintain the package structural integrity. This paper analyzes different storage arrays for a typical 9977 packaging for thermal considerations and makes recommendations for their safe storage under normal operating conditions.

  16. Motor Packaging with Consideration of Electromagnetic and Material...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    and Peer Evaluation ape035millerp.pdf More Documents & Publications Alnico and Ferrite Hybrid Excitation Electric Machines Motor Packaging with Consideration of...

  17. ESTIMATION OF RADIOLYTIC GAS GENERATION RATE FOR CYLINDRICAL RADIOACTIVE WASTE PACKAGES - APPLICATION TO SPENT ION EXCHANGE RESIN CONTAINERS

    SciTech Connect (OSTI)

    Husain, A.; Lewis, Brent J.

    2003-02-27T23:59:59.000Z

    Radioactive waste packages containing water and/or organic substances have the potential to radiolytically generate hydrogen and other combustible gases. Typically, the radiolytic gas generation rate is estimated from the energy deposition rate and the radiolytic gas yield. Estimation of the energy deposition rate must take into account the contributions from all radionuclides. While the contributions from non-gamma emitting radionuclides are relatively easy to estimate, an average geometry factor must be computed to determine the contribution from gamma emitters. Hitherto, no satisfactory method existed for estimating the geometry factors for a cylindrical package. In the present study, a formulation was developed taking into account the effect of photon buildup. A prototype code, called PC-CAGE, was developed to numerically solve the integrals involved. Based on the selected dimensions for a cylinder, the specified waste material, the photon energy of interest and a value for either the absorption or attenuation coefficient, the code outputs values for point and average geometry factors. These can then be used to estimate the internal dose rate to the material in the cylinder and hence to calculate the radiolytic gas generation rate. Besides the ability to estimate the rates of radiolytic gas generation, PC-CAGE can also estimate the dose received by the container material. This is based on values for the point geometry factors at the surface of the cylinder. PC-CAGE was used to calculate geometry factors for a number of cylindrical geometries. Estimates for the absorbed dose rate in container material were also obtained. The results for Ontario Power Generation's 3 m3 resin containers indicate that about 80% of the source gamma energy is deposited internally. In general, the fraction of gamma energy deposited internally depends on the dimensions of the cylinder, the material within it and the photon energy; the fraction deposited increases with increasing dimensions of the cylinder and decreases with increasing photon energy.

  18. AGING PERFORMANCE OF VITON GLT O-RINGS IN RADIOACTIVE MATERIAL PACKAGES

    SciTech Connect (OSTI)

    Skidmore, E; Kerry Dunn, K; Elizabeth Hoffman, E; Elise Fox, E; Kathryn Counts, K

    2007-05-07T23:59:59.000Z

    Radioactive material packages used for transportation of plutonium-bearing materials often contain multiple O-ring seals for containment. Packages such as the Model 9975 are also being used for interim storage of Pu-bearing materials at the Savannah River Site (SRS). One of the seal materials used in such packages is Viton{reg_sign} GLT fluoroelastomer. The aging behavior of containment vessel O-rings based on Viton{reg_sign} GLT at long-term containment term storage conditions is being characterized to assess its performance in such applications. This paper summarizes the program and test results to date.

  19. Shipment of Small Quantities of Unspecified Radioactive Material in Chalfant Packagings

    SciTech Connect (OSTI)

    Smith, Allen; Abramczyk, Glenn; Nathan, Steven; Bellamy, Steve

    2009-06-12T23:59:59.000Z

    In the post 6M era, radioactive materials package users are faced with the disciplined operations associated with use of Certified Type B packagings. Many DOE, commercial and academic programs have a requirement to ship and/or store small masses of poorly characterized or unspecified radioactive material. For quantities which are small enough to be fissile exempt and have low radiation levels, the materials could be transported in a package which provides the required containment level. Because their Chalfant type containment vessels meet the highest standard of containment (helium leak-tight), the 9975, 9977, and 9978 are capable of transporting any of these contents. The issues associated with certification of a high-integrity, general purpose package for shipping small quantities of unspecified radioactive material are discussed and certification of the packages for this mission is recommended.

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

    SciTech Connect (OSTI)

    Davis, M.S. (ed.)

    1985-12-01T23:59:59.000Z

    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.

  1. The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

    SciTech Connect (OSTI)

    Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

    2003-09-15T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Soo, P. (ed.)

    1986-01-01T23:59:59.000Z

    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.

  3. Compilation of current literature on seals, closures, and leakage for radioactive material packagings

    SciTech Connect (OSTI)

    Warrant, M.M.; Ottinger, C.A.

    1989-01-01T23:59:59.000Z

    This report presents an overview of the features that affect the sealing capability of radioactive material packagings currently certified by the US Nuclear Regulatory Commission. The report is based on a review of current literature on seals, closures, and leakage for radioactive material packagings. Federal regulations that relate to the sealing capability of radioactive material packagings, as well as basic equations for leakage calculations and some of the available leakage test procedures are presented. The factors which affect the sealing capability of a closure, including the properties of the sealing surfaces, the gasket material, the closure method and the contents are discussed in qualitative terms. Information on the general properties of both elastomer and metal gasket materials and some specific designs are presented. A summary of the seal material, closure method, and leakage tests for currently certified packagings with large diameter seals is provided. 18 figs., 9 tabs.

  4. Packaging and Transfer of Materials of National Security Interest Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-09-29T23:59:59.000Z

    The purpose of this Technical Manual is to establish requirements for operational safety controls for onsite operations. This Technical Manual provides Department of Energy (DOE) technical safety requirements and policy objectives for development of an onsite packaging and transfer program, pursuant to DOE O 461.1; the DOE contractor must document this program in its onsite packaging and transfer manual/procedures. Does not cancel other directives.

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

    SciTech Connect (OSTI)

    Farmer, J.C.; McCright, R.D.

    2000-01-28T23:59:59.000Z

    Alloy 22 is an extremely Corrosion Resistant Material, with a very stable passive film. Based upon exposures in the LTCTF, the GC rates of Alloy 22 are typically below the level of detection, with four outliers having reported rates up to 0.75 #mu#m per year. In any event, over the 10,000 year life of the repository, GC of the Alloy 22 (assumed to be 2 cm thick) should not be life limiting. Because measured corrosion potentials are far below threshold potentials, localized breakdown of the passive film is unlikely under plausible conditions, even in SSW at 120 deg C. The pH in ambient-temperature crevices formed from Alloy 22 have been determined experimentally, with only modest lowering of the crevice pH observed under plausible conditions. Extreme lowering of the crevice pH was only observed under situations where the applied potential at the crevice mouth was sufficient to result in catastrophic breakdown of the passive film above the threshold potential in non-buffered conditions not characteristic of the Yucca Mountain environment. In cases where naturally ocurring buffers are present in the crevice solution, little or no lowering of the pH was observed, even with significant applied potential. With exposures of twelve months, no evidence of crevice corrosion has been observed in SDW, SCW and SAW at temperatures up to 90 deg C. An abstracted model has been presented, with parameters determined experimentally, that should enable performance assessment to account for the general and localized corrosion of this material. A feature of this model is the use of the materials specification to limit the range of corrosion and threshold potentials, thereby making sure that substandard materials prone to localized attack are avoided. Model validation will be covered in part by a companion SMR on abstraction of this model.

  6. RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

    SciTech Connect (OSTI)

    KOZLOWSKI, S.D.

    2007-05-30T23:59:59.000Z

    This technical basis document was developed to support RPP-23429, Preliminary Documented Safety Analysis for the Demonstration Bulk Vitrification System (PDSA) and RPP-23479, Preliminary Documented Safety Analysis for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Facility. The main document describes the risk binning process and the technical basis for assigning risk bins to the representative accidents involving the release of dried radioactive waste materials from the Demonstration Bulk Vitrification System (DBVS) and to the associated represented hazardous conditions. Appendices D through F provide the technical basis for assigning risk bins to the representative dried waste release accident and associated represented hazardous conditions for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Packaging Unit (WPU). The risk binning process uses an evaluation of the frequency and consequence of a given representative accident or represented hazardous condition to determine the need for safety structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls. A representative accident or a represented hazardous condition is assigned to a risk bin based on the potential radiological and toxicological consequences to the public and the collocated worker. Note that the risk binning process is not applied to facility workers because credible hazardous conditions with the potential for significant facility worker consequences are considered for safety-significant SSCs and/or TSR-level controls regardless of their estimated frequency. The controls for protection of the facility workers are described in RPP-23429 and RPP-23479. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described below.

  7. Directory of certificates of compliance for radioactive materials packages

    SciTech Connect (OSTI)

    NONE

    1997-10-01T23:59:59.000Z

    The purpose of this directory is to make available a convenient source of information on packagings approved by the U.S. Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volumes 1 and 2. An alphabetical listing by user name is included in the back of Volume 3 for approved Quality Assurance programs. The reports include a listing of all users of each package design and approved Quality Assurance programs prior to the publication date of the directory. Comments to make future revisions of this directory more useful are invited and should be directed to the Spent Fuel Project Office, U.S. Nuclear Regulatory Commission.

  8. COMPACTION OF FIBERBOARD OVERPACK MATERIALS IN A 9975 SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Stefek, T.; Daugherty, W.; Estochen, E.; Murphy, J.

    2010-05-27T23:59:59.000Z

    Compaction of lower layers in the 9975 fiberboard overpack has been observed in packages that contain excess moisture. Dynamic loading of the package during transportation may also contribute to compaction of the fiberboard. This condition is being tested and analyzed to better understand these compaction mechanisms and provide a basis from which to evaluate their impact to the safety basis for transportation (Safety Analysis Report for Packaging) and storage (facility Design Safety Analysis) at the Savannah River Site (SRS). A test program has been developed and is being implemented to identify the extent of the compaction as a function of fiberboard moisture and typical transport dynamic loadings. Test conditions will be compared to regulatory requirements for dynamic loading. Characterization of the recovery of short-term compaction following the application of dynamic loading is also being evaluated. Interim results from this test program will be summarized.

  9. DRAFT - DOE O 460.1D, Hazardous Materials Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    The Order establishes safety requirements for the proper packaging and transportation of Department of offsite shipments and onsite transfers of radioactive and other hazardous materials, and for modal transportation.

  10. Packaging strategies for printed circuit board components. Volume I, materials & thermal stresses.

    SciTech Connect (OSTI)

    Neilsen, Michael K. (Kansas City Plant, Kansas City, MO); Austin, Kevin N.; Adolf, Douglas Brian; Spangler, Scott W.; Neidigk, Matthew Aaron; Chambers, Robert S.

    2011-09-01T23:59:59.000Z

    Decisions on material selections for electronics packaging can be quite complicated by the need to balance the criteria to withstand severe impacts yet survive deep thermal cycles intact. Many times, material choices are based on historical precedence perhaps ignorant of whether those initial choices were carefully investigated or whether the requirements on the new component match those of previous units. The goal of this program focuses on developing both increased intuition for generic packaging guidelines and computational methodologies for optimizing packaging in specific components. Initial efforts centered on characterization of classes of materials common to packaging strategies and computational analyses of stresses generated during thermal cycling to identify strengths and weaknesses of various material choices. Future studies will analyze the same example problems incorporating the effects of curing stresses as needed and analyzing dynamic loadings to compare trends with the quasi-static conclusions.

  11. Implementation Guide for Use with DOE O 460.2 Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-11-15T23:59:59.000Z

    The purpose of this guide is to assist those responsible for transporting and packaging Department materials, and to provide an understanding of Department policies on activities which supplement regulatory requirements.

  12. LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE

    SciTech Connect (OSTI)

    Dunn, K. [Savannah River National Laboratory; Bellamy, S. [Savannah River National Laboratory; Daugherty, W. [Savannah River National Laboratory; Sindelar, R. [Savannah River National Laboratory; Skidmore, E. [Savannah River National Laboratory

    2013-08-18T23:59:59.000Z

    Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintain integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.

  13. DOE nuclear material packaging manual: storage container requirements for plutonium oxide materials

    SciTech Connect (OSTI)

    Veirs, D Kirk [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    Loss of containment of nuclear material stored in containers such as food-pack cans, paint cans, or taped slip lid cans has generated concern about packaging requirements for interim storage of nuclear materials in working facilities such as the plutonium facility at Los Alamos National Laboratory (LANL). In response, DOE has recently issued DOE M 441.1 'Nuclear Material Packaging Manual' with encouragement from the Defense Nuclear Facilities Safety Board. A unique feature compared to transportation containers is the allowance of filters to vent flammable gases during storage. Defining commonly used concepts such as maximum allowable working pressure and He leak rate criteria become problematic when considering vented containers. Los Alamos has developed a set of container requirements that are in compliance with 441.1 based upon the activity of heat-source plutonium (90% Pu-238) oxide, which bounds the requirements for weapons-grade plutonium oxide. The pre and post drop-test He leak rates depend upon container size as well as the material contents. For containers that are routinely handled, ease of handling and weight are a major consideration. Relatively thin-walled containers with flat bottoms are desired yet they cannot be He leak tested at a differential pressure of one atmosphere due to the potential for plastic deformation of the flat bottom during testing. The He leak rates and He leak testing configuration for containers designed for plutonium bearing materials will be presented. The approach to meeting the other manual requirements such as corrosion and thermal degradation resistance will be addressed. The information presented can be used by other sites to evaluate if their conditions are bounded by LANL requirements when considering procurement of 441.1 compliant containers.

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

    SciTech Connect (OSTI)

    LOCKREM, L.L.

    1999-08-13T23:59:59.000Z

    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.

  15. Material Recovery and Waste Form Development FY 2014 Accomplishments Report

    SciTech Connect (OSTI)

    Lori Braase

    2014-11-01T23:59:59.000Z

    Develop advanced nuclear fuel cycle separation and waste management technologies that improve current fuel cycle performance and enable a sustainable fuel cycle, with minimal processing, waste generation, and potential for material diversion.

  16. Life and stability testing of packaged low-cost energy storage materials

    SciTech Connect (OSTI)

    Frysinger, G.R.

    1980-07-01T23:59:59.000Z

    A low-cost laminated plastic film which is used to contain a Glauber's salt-based phase change thermal energy storage material in sausage-like containers called Chubs is discussed. The results of tests performed on the Chub packages themselves and on the thermal energy storage capacity of the packaged phase change material are described. From the test results, a set of specifications have been drawn up for a film material which will satisfactorily contain the phase change material under anticipated operating conditions. Calorimetric testing of the phase change material with thermal cycling indicates that a design capacity of 45 to 50 Btu/lb for a ..delta..T of 30/sup 0/F can be used for the packaged material.

  17. Molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

    Brummond, William A. (Livermore, CA); Upadhye, Ravindra S. (Pleasanton, CA); Pruneda, Cesar O. (Livermore, CA)

    1995-01-01T23:59:59.000Z

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor.

  18. Molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

    Brummond, W.A.; Upadhye, R.S.; Pruneda, C.O.

    1995-07-18T23:59:59.000Z

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor. 4 figs.

  19. INVESTIGATION OF THE PRESENCE OF DRUGSTORE BEETLES WITHIN CELOTEX ASSEMBLIES IN RADIOACTIVE MATERIAL PACKAGINGS

    SciTech Connect (OSTI)

    Loftin, B; Glenn Abramczyk, G

    2008-06-04T23:59:59.000Z

    During normal operations at the Department of Energy's Hanford Site in Hanford, WA, drugstore beetles, (Stegobium paniceum (L.) Coleoptera: Anobiidae), were found within the fiberboard subassemblies of two 9975 Shipping Packages. Initial indications were that the beetles were feeding on the Celotex{trademark} assemblies within the package. Celotex{trademark} fiberboard is used in numerous radioactive material packages serving as both a thermal insulator and an impact absorber for both normal conditions of transport and hypothetical accident conditions. The Department of Energy's Packaging Certification Program (EM-63) directed a thorough investigation to determine if the drugstore beetles were causing damage that would be detrimental to the safety performance of the Celotex{trademark}. The Savannah River National Laboratory is conducting the investigation with entomological expertise provided by Clemson University. The two empty 9975 shipping packages were transferred to the Savannah River National Laboratory in the fall of 2007. This paper will provide details and results of the ongoing investigation.

  20. Development on inelastic analysis acceptance criteria for radioactive material transportation packages

    SciTech Connect (OSTI)

    Ammerman, D.J.; Ludwigsen, J.S.

    1995-12-31T23:59:59.000Z

    The response of radioactive material transportation packages to mechanical accident loadings can be more accurately characterized by non-linear dynamic analysis than by the ``Equivalent dynamic`` static elastic analysis typically used in the design of these packages. This more accurate characterization of the response can lead to improved package safety and design efficiency. For non-linear dynamic analysis to become the preferred method of package design analysis, an acceptance criterion must be established that achieves an equivalent level of safety as the currently used criterion defined in NRC Regulatory Guide 7.6 (NRC 1978). Sandia National Laboratories has been conducting a study of possible acceptance criteria to meet this requirement. In this paper non-linear dynamic analysis acceptance criteria based on stress, strain, and strain-energy-density will be discussed. An example package design will be compared for each of the design criteria, including the approach of NRC Regulatory Guide 7.6.

  1. Teaching Old Packaging New Tricks - 12593

    SciTech Connect (OSTI)

    England, Jeffery L. [Savannah River National Laboratory, Aiken, South Carolina (United States); Shuler, James M. [U.S. Department of Energy, Packaging Certification Program, Germantown, MD (United States)

    2012-07-01T23:59:59.000Z

    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)

  2. FINAL REPORT WASTE PACKAGE MATERIALS PERFORMANCE PEER REVIEW...

    Broader source: Energy.gov (indexed) [DOE]

    J., and Szklarska-Smialowska, S., Corrosion Science, 15, 493 (1975). Martell A. E. and Smith R. M., 1998, Critically Selected Stability Constants of Metal Complexes Database...

  3. Removal of radioactive and other hazardous material from fluid waste

    DOE Patents [OSTI]

    Tranter, Troy J. (Idaho Falls, ID); Knecht, Dieter A. (Idaho Falls, ID); Todd, Terry A. (Aberdeen, ID); Burchfield, Larry A. (W. Richland, WA); Anshits, Alexander G. (Krasnoyarsk, RU); Vereshchagina, Tatiana (Krasnoyarsk, RU); Tretyakov, Alexander A. (Zheleznogorsk, RU); Aloy, Albert S. (St. Petersburg, RU); Sapozhnikova, Natalia V. (St. Petersburg, RU)

    2006-10-03T23:59:59.000Z

    Hollow glass microspheres obtained from fly ash (cenospheres) are impregnated with extractants/ion-exchangers and used to remove hazardous material from fluid waste. In a preferred embodiment the microsphere material is loaded with ammonium molybdophosphonate (AMP) and used to remove radioactive ions, such as cesium-137, from acidic liquid wastes. In another preferred embodiment, the microsphere material is loaded with octyl(phenyl)-N-N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) and used to remove americium and plutonium from acidic liquid wastes.

  4. Regulatory compliance in the design of packages used to transport radioactive materials

    SciTech Connect (OSTI)

    Raske, D.T.

    1993-06-01T23:59:59.000Z

    Shipments of radioactive materials within the regulatory jurisdiction of the US Department of Energy (DOE) must meet the package design requirements contained in Title 10 of the Code of Federal Regulations, Part 71, and DOE Order 5480.3. These regulations do not provide design criteria requirements, but only detail the approval standards, structural performance criteria, and package integrity requirements that must be met during transport. The DOE recommended design criterion for high-level Category I radioactive packagings is Section III, Division 1, of the ASME Boiler and Pressure Vessel Code. However, alternative design criteria may be used if all the design requirements are satisfied. The purpose of this paper is to review alternatives to the Code criteria and discuss their applicability to the design of containment vessels in packages for high-level radioactive materials. Issues such as design qualification by physical testing, the use of scale models, and problems encountered using a non-ASME design approach are addressed.

  5. Material selection for Multi-Function Waste Tank Facility tanks

    SciTech Connect (OSTI)

    Larrick, A.P.; Blackburn, L.D.; Brehm, W.F.; Carlos, W.C.; Hauptmann, J.P. [Westinghouse Hanford Co., Richland, WA (United States); Danielson, M.J.; Westerman, R.E. [Pacific Northwest Lab., Richland, WA (United States); Divine, J.R. [ChemMet Ltd., West Richland, WA (United States); Foster, G.M. [ICF Kaiser Hanford Co., Richland, WA (United States)

    1995-03-01T23:59:59.000Z

    This paper briefly summarizes the history of the materials selection for the US Department of Energy`s high-level waste carbon steel storage tanks. It also provides an evaluation of the materials for the construction of new tanks at the evaluation of the materials for the construction of new tanks at the Multi-Function Waste Tank Facility. The evaluation included a materials matrix that summarized the critical design, fabrication, construction, and corrosion resistance requirements: assessed. each requirement: and cataloged the advantages and disadvantages of each material. This evaluation is based on the mission of the Multi-Function Waste Tank Facility. On the basis of the compositions of the wastes stored in Hanford waste tanks, it is recommended that tanks for the Multi-Function Waste Tank Facility be constructed of ASME SA 515, Grade 70, carbon steel.

  6. FEASIBILITY OF TARGET MATERIAL RECYCLING AS WASTE MANAGEMENT ALTERNATIVE

    E-Print Network [OSTI]

    California at San Diego, University of

    FEASIBILITY OF TARGET MATERIAL RECYCLING AS WASTE MANAGEMENT ALTERNATIVE L. EL-GUEBALY,* P. WILSON for Publication February 3, 2004 The issue of waste management has been studied simultaneously along with the development of the ARIES heavy-ion-driven inertial fusion energy (IFE) concept. Options for waste management

  7. Diffusion and Leaching Behavior of Radionuclides in Category 3 Waste Encasement Concrete and Soil Fill Material – Summary Report

    SciTech Connect (OSTI)

    Mattigod, Shas V.; Wellman, Dawn M.; Bovaird, Chase C.; Parker, Kent E.; Clayton, Libby N.; Powers, Laura; Recknagle, Kurtis P.; Wood, Marcus I.

    2011-08-31T23:59:59.000Z

    One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Such concrete encasement would contain and isolate the waste packages from the hydrologic environment and would act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed, and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The mobilized radionuclides may escape from the encased concrete by mass flow and/or diffusion and move into the surrounding subsurface environment. Therefore, it is necessary to assess the performance of the concrete encasement structure and the ability of the surrounding soil to retard radionuclide migration. The retardation factors for radionuclides contained in the waste packages can be determined from measurements of diffusion coefficients for these contaminants through concrete and fill material. Some of the mobilization scenarios include (1) potential leaching of waste form before permanent closure cover is installed; (2) after the cover installation, long-term diffusion of radionuclides from concrete waste form into surrounding fill material; (3) diffusion of radionuclides from contaminated soils into adjoining concrete encasement and clean fill material. Additionally, the rate of diffusion of radionuclides may be affected by the formation of structural cracks in concrete, the carbonation of the buried waste form, and any potential effect of metallic iron (in the form of rebars) on the mobility of radionuclides. The radionuclides iodine-129 ({sup 129}I), technetium-99 ({sup 99}Tc), and uranium-238 ({sup 238}U) are identified as long-term dose contributors in Category 3 waste (Mann et al. 2001; Wood et al. 1995). Because of their anionic nature in aqueous solutions, {sup 129}I, {sup 99}Tc, and carbonate-complexed {sup 238}U may readily leach into the subsurface environment (Serne et al. 1989, 1992a, b, 1993, and 1995). The leachability and/or diffusion of radionuclide species must be measured to assess the long-term performance of waste grouts when contacted with vadose-zone pore water or groundwater. Although significant research has been conducted on the design and performance of cementitious waste forms, the current protocol conducted to assess radionuclide stability within these waste forms has been limited to the Toxicity Characteristic Leaching Procedure, Method 1311 Federal Registry (EPA 1992) and ANSI/ANS-16.1 leach test (ANSI 1986). These tests evaluate the performance under water-saturated conditions and do not evaluate the performance of cementitious waste forms within the context of waste repositories which are located within water-deficient vadose zones. Moreover, these tests assess only the diffusion of radionuclides from concrete waste forms and neglect evaluating the mechanisms of retention, stability of the waste form, and formation of secondary phases during weathering, which may serve as long-term secondary hosts for immobilization of radionuclides. The results of recent investigations conducted under arid and semi-arid conditions (Al-Khayat et al. 2002; Garrabrants et al. 2002; Garrabrants and Kosson 2003; Garrabrants et al. 2004; Gervais et al. 2004; Sanchez et al. 2002; Sanchez et al. 2003) provide valuable information suggesting structural and chemical changes to concrete waste forms which may affect contaminant containm

  8. Packaging and Transfer or Transportation of Materials of National Security Interest

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-09-29T23:59:59.000Z

    To establish requirements and responsibilities for the Transportation Safeguards System (TSS) packaging and transportation and onsite transfer of nuclear explosives, nuclear components, Naval nuclear fuel elements, Category I and Category II special nuclear materials, special assemblies, and other materials of national security interest. Cancels: DOE 5610.12 and DOE 5610.14.

  9. System for chemically digesting low level radioactive, solid waste material

    DOE Patents [OSTI]

    Cowan, Richard G. (Kennewick, WA); Blasewitz, Albert G. (Richland, WA)

    1982-01-01T23:59:59.000Z

    An improved method and system for chemically digesting low level radioactive, solid waste material having a high through-put. The solid waste material is added to an annular vessel (10) substantially filled with concentrated sulfuric acid. Concentrated nitric acid or nitrogen dioxide is added to the sulfuric acid within the annular vessel while the sulfuric acid is reacting with the solid waste. The solid waste is mixed within the sulfuric acid so that the solid waste is substantilly fully immersed during the reaction. The off gas from the reaction and the products slurry residue is removed from the vessel during the reaction.

  10. Directory of certificiates of compliance for radioactive materials packages: Report of NRC approved packages. Revision 19, Volume 1

    SciTech Connect (OSTI)

    NONE

    1996-10-01T23:59:59.000Z

    This directory provides information on packagings approved by the U.S. Nuclear Regulatory Commission.

  11. Determination of Fire Enviroment in Stacked Cargo Containers with Radioactive Materials Packages

    SciTech Connect (OSTI)

    Arviso, M.; Bobbe, J.G.; Dukart, R.D.; Koski, J.A.

    1999-05-01T23:59:59.000Z

    Results from a Fire Test with a three-by-three stack of standard 6 m long International Standards Organization shipping containers containing combustible fuels and empty radioactive materials packages are reported and discussed. The stack is intended to simulate fire conditions that could occur during on-deck stowage on container cargo ships. The fire is initated by locating the container stack adjacent to a 9.8 x 6 m pool fire. Temperatures of both cargoes (empty and simulated radioactive materials packages) and containers are recorded and reported. Observations on the duration, intensity and spread of the fire are discussed. Based on the results, models for simulation of fire exposure of radioactive materials packages in such fires are suggested.

  12. Material and energy recovery in integrated waste management systems: Project overview and main results

    SciTech Connect (OSTI)

    Consonni, Stefano, E-mail: stefano.consonni@polimi.it [Department of Energy, Politecnico di Milano, Via Lambruschini 4, 20156 Milan (Italy); Giugliano, Michele [DIIAR, Environmental Section, Politecnico di Milano, P.za L. Da Vinci 32, 20133 Milan (Italy); Massarutto, Antonio [Dse, Universita degli Studi di Udine and IEFE, Via Tomadini 30/a, 33100 Udine (Italy); Ragazzi, Marco [Department of Civil and Environmental Engineering, University of Trento, Via Mesiano 77, 38123 Trento (Italy); Saccani, Cesare [DIEM, University of Bologna, Viale Risorgimento 2, 40136 Bologna (Italy)

    2011-09-15T23:59:59.000Z

    Highlights: > The source separation level (SSL) of waste management system does not qualify adequately the system. > Separately collecting organic waste gives less advantages than packaging materials. > Recycling packaging materials (metals, glass, plastics, paper) is always attractive. > Composting and anaerobic digestion of organic waste gives questionable outcomes. > The critical threshold of optimal recycling seems to be a SSL of 50%. - Abstract: This paper describes the context, the basic assumptions and the main findings of a joint research project aimed at identifying the optimal breakdown between material recovery and energy recovery from municipal solid waste (MSW) in the framework of integrated waste management systems (IWMS). The project was carried out from 2007 to 2009 by five research groups at Politecnico di Milano, the Universities of Bologna and Trento, and the Bocconi University (Milan), with funding from the Italian Ministry of Education, University and Research (MIUR). Since the optimization of IWMSs by analytical methods is practically impossible, the search for the most attractive strategy was carried out by comparing a number of relevant recovery paths from the point of view of mass and energy flows, technological features, environmental impact and economics. The main focus has been on mature processes applicable to MSW in Italy and Europe. Results show that, contrary to a rather widespread opinion, increasing the source separation level (SSL) has a very marginal effects on energy efficiency. What does generate very significant variations in energy efficiency is scale, i.e. the size of the waste-to-energy (WTE) plant. The mere value of SSL is inadequate to qualify the recovery system. The energy and environmental outcome of recovery depends not only on 'how much' source separation is carried out, but rather on 'how' a given SSL is reached.

  13. 9975 SHIPPING PACKAGE PERFORMANCE OF ALTERNATE MATERIALS FOR LONG-TERM STORAGE APPLICATION

    SciTech Connect (OSTI)

    Skidmore, E.; Hoffman, E.; Daugherty, W.

    2010-02-24T23:59:59.000Z

    The Model 9975 shipping package specifies the materials of construction for its various components. With the loss of availability of material for two components (cane fiberboard overpack and Viton{reg_sign} GLT O-rings), alternate materials of construction were identified and approved for use for transport (softwood fiberboard and Viton{reg_sign} GLT-S O-rings). As these shipping packages are part of a long-term storage configuration at the Savannah River Site, additional testing is in progress to verify satisfactory long-term performance of the alternate materials under storage conditions. The test results to date can be compared to comparable results on the original materials of construction to draw preliminary conclusions on the performance of the replacement materials.

  14. Yucca Mountain project canister material corrosion studies as applied to the electrometallurgical treatment metallic waste form

    SciTech Connect (OSTI)

    Keiser, D.D.

    1996-11-01T23:59:59.000Z

    Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne`s waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne`s metal waste form in light of the Yucca Mountain activities.

  15. assay package insert: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    disposal. An important task for waste management organizations is to translate general waste acceptance requirements into detailed waste package specifications. Waste package...

  16. Development and evaluation of measurement devices used to support testing of radioactive material transportation packages

    SciTech Connect (OSTI)

    Uncapher, W. L.; Ammerman, D. J.; Stenberg, D.R.; Bronowski, D. R.; Arviso, M.

    1992-01-01T23:59:59.000Z

    Radioactive material package designers use structural testing to verify and demonstrate package performance. A major part of evaluating structural response is the collection of instrumentation measurement data. Sandia National Laboratories (SNL) has an ongoing program to develop and evaluate measurement devices to support testing of radioactive material packages. Measurement devices developed in support of this activity include evaluation channels, ruggedly constructed linear variable differential transformers, and piezoresistive accelerometers with enhanced measurement capabilities. In addition to developing measurement devices, a method has been derived to evaluate accelerometers and strain gages for measurement repeatability, ruggedness, and manufacturers' calibration data under both laboratory and field conditions. The developed measurement devices and evaluation technique will be discussed and the results of the evaluation will be presented.

  17. Use of inelastic design for radioactive material transportation packages

    SciTech Connect (OSTI)

    Heinstein, M.W.; Ammerman, D.J.

    1993-12-01T23:59:59.000Z

    There is much interest within the radioactive material transportation container design community in the use of inelastic analysis. In other industries where inelastic analysis is used in design there is typically an improved knowledge of the capacity of the structure and a more efficient use of material. This report describes the results of a program in which the incentives for inelastic analysis for radioactive material transport container design were investigated to determine if there are similar benefits. Detailed are the elastic and inelastic analyses of two containers subjected to impacts onto a rigid target following a thirty-foot free fall in end-on, side-on, and center-of-gravity- over-corner orientations.

  18. Cleanup Verification Package for the 618-2 Burial Ground

    SciTech Connect (OSTI)

    W. S. Thompson

    2006-12-28T23:59:59.000Z

    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.

  19. Organic and inorganic hazardous waste stabilization utilizing fossil fuel combustion waste materials

    SciTech Connect (OSTI)

    Netzel, D.A.; Lane, D.C.; Brown, M.A.; Raska, K.A.; Clark, J.A.; Rovani, J.F.

    1993-09-01T23:59:59.000Z

    A laboratory study was conducted at the Western Research Institute to evaluate the ability of innovative clean coal technology (ICCT) waste to stabilize organic and inorganic constituents of hazardous wastes. The four ICCT wastes used in this study were: (1) the Tennessee Valley Authority (TVA) atmospheric fluidized bed combustor (AFBC) waste, (2) the TVA spray dryer waste, (3) the Laramie River Station spray dryer waste, and (4) the Colorado-Ute AFBC waste. Four types of hazardous waste stream materials were obtained and chemically characterized for use in evaluating the ability of the ICCT wastes to stabilize hazardous organic and inorganic wastes. The wastes included an API separator sludge, mixed metal oxide-hydroxide waste, metal-plating sludge, and creosote-contaminated soil. The API separator sludge and creosote-contaminated soil are US Environmental Protection Agency (EPA)-listed hazardous wastes and contain organic contaminants. The mixed metal oxide-hydroxide waste and metal-plating sludge (also an EPA-listed waste) contain high concentrations of heavy metals. The mixed metal oxide-hydroxide waste fails the Toxicity Characteristic Leaching Procedure (TCLP) for cadmium, and the metal-plating sludge fails the TCLP for chromium. To evaluate the ability of the ICCT wastes to stabilize the hazardous wastes, mixtures involving varying amounts of each of the ICCT wastes with each of the hazardous wastes were prepared, allowed to equilibrate, and then leached with deionized, distilled water. The leachates were analyzed for the hazardous constituent(s) of interest using the Toxicity Characteristic Leaching Procedure.

  20. Test plan for buried waste containment system materials

    SciTech Connect (OSTI)

    Weidner, J.; Shaw, P.

    1997-03-01T23:59:59.000Z

    The objectives of the FY 1997 barrier material work at the Idaho National Engineering and Environmental Laboratory are to (1) select a waste barrier material and verify that it is compatible with the Buried Waste Containment System Process, and (2) determine if, and how, the Buried Waste Containment System emplacement process affects the material properties and performance (on proof of principle scale). This test plan describes a set of measurements and procedures used to validate a waste barrier material for the Buried Waste Containment System. A latex modified proprietary cement manufactured by CTS Cement Manufacturing Company will be tested. Emplacement properties required for the Buried Waste Containment System process are: slump between 8 and 10 in., set time between 15 and 30 minutes, compressive strength at set of 20 psi minimum, and set temperature less than 100{degrees}C. Durability properties include resistance to degradation from carbonate, sulfate, and waste-site soil leachates. A set of baseline barrier material properties will be determined to provide a data base for comparison with the barrier materials when tested in the field. The measurements include permeability, petrographic analysis to determine separation and/or segregation of mix components, and a set of mechanical properties. The measurements will be repeated on specimens from the field test material. The data will be used to determine if the Buried Waste Containment System equipment changes the material. The emplacement properties will be determined using standard laboratory procedures and instruments. Durability of the barrier material will be evaluated by determining the effect of carbonate, sulfate, and waste-site soil leachates on the compressive strength of the barrier material. The baseline properties will be determined using standard ASTM procedures. 9 refs., 1 fig., 2 tabs.

  1. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    SciTech Connect (OSTI)

    H. Wang

    1997-01-23T23:59:59.000Z

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  2. DEVELOPMENT OF BURN TEST SPECIFICATIONS FOR FIRE PROTECTION MATERIALS IN RAM PACKAGES

    SciTech Connect (OSTI)

    Gupta, N.

    2010-03-03T23:59:59.000Z

    The regulations in 10 CFR 71 require that the radioactive material (RAM) packages must be able to withstand specific fire conditions given in 10 CFR 71.73 during Hypothetical Accident Conditions (HAC). This requirement is normally satisfied by extensive testing of full scale test specimens under required test conditions. Since fire test planning and execution is expensive and only provides a single snapshot into a package performance, every effort is made to minimize testing and supplement tests with results from computational thermal models. However, the accuracy of such thermal models depends heavily on the thermal properties of the fire insulating materials that are rarely available at the regulatory fire temperatures. To the best of authors knowledge no test standards exist that could be used to test the insulating materials and derive their thermal properties for the RAM package design. This paper presents a review of the existing industry fire testing standards and proposes testing methods that could serve as a standardized specification for testing fire insulating materials for use in RAM packages.

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

    SciTech Connect (OSTI)

    Tyacke, M.

    1993-08-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Trennel, A.J.

    1995-12-31T23:59:59.000Z

    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. Corrosion Protection Provided by PV Module Packaging Materials

    SciTech Connect (OSTI)

    Jorgensen, G. J.; Kempe, M. D.; Terwilliger, K. M.; McMahon, T. J.

    2005-11-01T23:59:59.000Z

    The ability of glass/glass and glass/breathable backsheet constructions laminated with various encapsulant and/or edge seal materials to protect thin-film aluminum coatings deposited onto glass substrates was assessed. Although they provide the best moisture barrier available, glass/glass laminate constructions can trap harmful compounds that catalyze moisture-driven corrosion of the aluminum. Constructions with breathable backsheets allow higher rates of moisture ingress, but also allow egress of deleterious substances that can result in decreased corrosion.

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

    SciTech Connect (OSTI)

    Marusich, Robert M.

    2012-01-25T23:59:59.000Z

    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).

  7. A COMPARISON OF TWO THERMAL INSULATION AND STRUCTURAL MATERIALS FOR USE IN TYPE B PACKAGINGS

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2010-07-16T23:59:59.000Z

    This paper presents the summary of design features and test results of two Type B Shipping Package prototype configurations comprising different insulating materials developed by the Savannah River National Laboratory (SRNL) for the Department of Energy. The materials evaluated, a closed-cell polyurethane foam and a vacuformed ceramic fiber material, were selected to provide adequate structural protection to the package containment vessel during Normal Conditions of Transport (NCT) and Hypothetical Accident Condition (HAC) events and to provide thermal protection during the HAC fire. Polyurethane foam has been used in shipping package designs for many years because of the stiffness it provides to the structure and because of the thermal protection it provides during fire scenarios. This comparison describes how ceramic fiber material offers an alternative to the polyurethane foam in a specific overpack design. Because of the high operating temperature ({approx}2,300 F) of the ceramic material, it allows for contents with higher heat loads to be shipped than is possible with polyurethane foam. Methods of manufacturing and design considerations using the two materials will be addressed.

  8. Radioisotope Thermoelectric Generator Package O-Ring Seal Material Validation Testing

    SciTech Connect (OSTI)

    Adkins, H.E.; Ferrell, P.C.; Knight, R.C.

    1994-09-30T23:59:59.000Z

    The Radioisotope Thermoelectric Generator Package O-Ring Seal Material Validation Test was conducted to validate the use of the Butyl material as a primary seal throughout the required temperature range. Three tests were performed at (1) 233 K ({minus}40 {degrees}F), (2) a specified operating temperature, and (3) 244 K ({minus}20 {degrees}F) before returning to room temperature. Helium leak tests were performed at each test point to determine seal performance. The two major test objectives were to establish that butyl rubber material would maintain its integrity under various conditions and within specified parameters and to evaluate changes in material properties.

  9. Safety evaluation for packaging 222-S laboratory cargo tank for onetime type B material shipment

    SciTech Connect (OSTI)

    Nguyen, P.M.

    1994-08-19T23:59:59.000Z

    The purpose of this Safety Evaluation for Packaging (SEP) is to evaluate and document the safety of the onetime shipment of bulk radioactive liquids in the 222-S Laboratory cargo tank (222-S cargo tank). The 222-S cargo tank is a US Department of Transportation (DOT) MC-312 specification (DOT 1989) cargo tank, vehicle registration number HO-64-04275, approved for low specific activity (LSA) shipments in accordance with the DOT Title 49, Code of Federal Regulations (CFR). In accordance with the US Department of Energy, Richland Operations Office (RL) Order 5480.1A, Chapter III (RL 1988), an equivalent degree of safety shall be provided for onsite shipments as would be afforded by the DOT shipping regulations for a radioactive material package. This document demonstrates that this packaging system meets the onsite transportation safety criteria for a onetime shipment of Type B contents.

  10. OFFICE WASTE DATA 2010 Recyclable Materials 1680 tons / 62%

    E-Print Network [OSTI]

    Guillas, Serge

    OFFICE WASTE DATA 2010 Recyclable Materials 1680 tons / 62% Landfill 1080 tons / 38% Electricals 36 Landfill As of Monday 7 March 2011, no general waste generated from the Bloomsbury Campus has been sent to landfill. Through partnership between UCL Estates and Office and General, an agreement has been reached

  11. Practical Thermal Evaluation Methods For HAC Fire Analysis In Type B Radiaoactive Material (RAM) Packages

    SciTech Connect (OSTI)

    Abramczyk, Glenn; Hensel, Stephen J; Gupta, Narendra K.

    2013-03-28T23:59:59.000Z

    Title 10 of the United States Code of Federal Regulations Part 71 for the Nuclear Regulatory Commission (10 CFR Part 71.73) requires that Type B radioactive material (RAM) packages satisfy certain Hypothetical Accident Conditions (HAC) thermal design requirements to ensure package safety during accidental fire conditions. Compliance with thermal design requirements can be met by prototype tests, analyses only or a combination of tests and analyses. Normally, it is impractical to meet all the HAC using tests only and the analytical methods are too complex due to the multi-physics non-linear nature of the fire event. Therefore, a combination of tests and thermal analyses methods using commercial heat transfer software are used to meet the necessary design requirements. The authors, along with his other colleagues at Savannah River National Laboratory in Aiken, SC, USA, have successfully used this 'tests and analyses' approach in the design and certification of several United States' DOE/NNSA certified packages, e.g. 9975, 9977, 9978, 9979, H1700, and Bulk Tritium Shipping Package (BTSP). This paper will describe these methods and it is hoped that the RAM Type B package designers and analysts can use them for their applications.

  12. APPLICATION FO FLOW FORMING FOR USE IN RADIOACTIVE MATERIAL PACKAGING DESIGNS

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.; Abramczyk, G.

    2012-07-11T23:59:59.000Z

    This paper reports on the development and testing performed to demonstrate the use of flow forming as an alternate method of manufacturing containment vessels for use in radioactive material shipping packaging designs. Additionally, ASME Boiler and Pressure Vessel Code, Section III, Subsection NB compliance along with the benefits compared to typical welding of containment vessels will be discussed. SRNL has completed fabrication development and the testing on flow formed containment vessels to demonstrate the use of flow forming as an alternate method of manufacturing a welded 6-inch diameter containment vessel currently used in the 9975 and 9977 radioactive material shipping packaging. Material testing and nondestructive evaluation of the flow formed parts demonstrate compliance to the minimum material requirements specified in applicable parts of ASME Boiler and Pressure Vessel Code, Section II. Destructive burst testing shows comparable results to that of a welded design. The benefits of flow forming as compared to typical welding of containment vessels are significant: dimensional control is improved due to no weld distortion; less final machining; weld fit-up issues associated with pipes and pipe caps are eliminated; post-weld non-destructive testing (i.e., radiography and die penetrant tests) is not necessary; and less fabrication steps are required. Results presented in this paper indicate some of the benefits in adapting flow forming to design of future radioactive material shipping packages containment vessels.

  13. Thermochemical conversion of waste materials to valuable products

    SciTech Connect (OSTI)

    Saraf, S. [Engineering Technologies, Lombard, IL (United States)

    1997-12-31T23:59:59.000Z

    The potential offered by a large variety of solid and liquid wastes for generating value added products is widely recognized. Extensive research and development has focused on developing technologies to recover energy and valuable products from waste materials. These treatment technologies include use of waste materials for direct combustion, upgrading the waste materials into useful fuel such as fuel gas or fuel oil, and conversion of waste materials into higher value products for the chemical industry. Thermal treatment in aerobic (with oxygen) conditions or direct combustion of waste materials in most cases results in generating air pollution and thereby requiring installation of expensive control devices. Thermochemical conversion in aerobic (without oxygen) conditions, referred to as thermal decomposition (destructive distillation) results in formation of usable liquid, solid, and gaseous products. Thermochemical conversion includes gasification, liquefaction, and thermal decomposition (pyrolysis). Each thermochemical conversion process yields a different range of products and this paper will discuss thermal decomposition in detail. This paper will also present results of a case study for recovering value added products, in the form of a liquid, solid, and gas, from thermal decomposition of waste oil and scrap tires. The product has a high concentration of benzene, xylene, and toluene. The solid product has significant amounts of carbon black and can be used as an asphalt modifier for road construction. The gas product is primarily composed of methane and is used for heating the reactor.

  14. Productivity Techniques and Quality Aspects in the Criticality Safety Evaluation of Y-12 Type-B Fissile Material Packages

    SciTech Connect (OSTI)

    DeClue, J. F.

    2011-06-28T23:59:59.000Z

    The inventory of certified Type-B fissile material packages consists of ten performance-based packages for offsite transportation purposes, serving transportation programs at the Y-12 National Security Complex. The containment vessels range from 5 to 19 in. in diameter and from 17 to 58 in. in height. The drum assembly external to the containment vessel ranges from 18 to 34 in. in diameter and from 26 to 71 in. in height. The weight of the packaging (drum assembly and containment vessel) ranges from 239 to 1550 lb. The older DT-nn series of Cellotex-based packages are being phased-out and replaced by a new generation of Kaolite-based ('Y-12 patented insulation') packages capable of withstanding the dynamic crush test 10 CFR 71.73(c)(2). Three replacement packages are in various stages of development; two are in use. The U.S. Department of Transportation (DOT) 6M specification package, which does not conform to the U.S. Nuclear Regulatory Commission requirements for Type-B packages, is no longer authorized for service on public roads. The ES-3100 shipping package is an example of a Kaolite-based Type-B fissile material package developed as a replacement package for the DOT 6M. With expanded utility, the ES-3100 is designed and licensed for transporting highly enriched uranium and plutonium materials on public roads. The ES-3100 provides added capability for air transport of up to 7-kg quantities of uranium material. This paper presents the productivity techniques and quality aspects in the criticality safety evaluation of Y-12 packages using the ES-3100 as an example.

  15. MATERIALS COMPATIBILITY OF SNAP FUEL COMPONENTS DURING SHIPMENT IN 9975 PACKAGING

    SciTech Connect (OSTI)

    Vormelker, P

    2006-11-14T23:59:59.000Z

    Materials Science and Technology has evaluated materials compatibility for the SNAP (Systems for Nuclear Auxiliary Power) fuel for containment within a 9975 packaging assembly for a shipping period of one year. The evaluation included consideration for potential for water within the convenience can, corrosion from water, galvanic corrosion, tape degradation, and thermal expansion risk. Based on a review of existing literature and assumed conditions, corrosion and/or degradation of the 304 stainless steel (SS) Primary Containment Vessel (PCV) and the 304 stainless steel convenience cans containing the SNAP fuel is not significant to cause failure during the 1 year time shipping period in the 9975 packaging assembly. However, storage beyond the 1 year shipping period has not been validated.

  16. Material and energy recovery in integrated waste management systems. An evaluation based on life cycle assessment

    SciTech Connect (OSTI)

    Giugliano, Michele; Cernuschi, Stefano [Politecnico di Milano - DIIAR, Environmental Section, P.zza Leonardo da Vinci, 32, 20133 Milano (Italy); Grosso, Mario, E-mail: mario.grosso@polimi.it [Politecnico di Milano - DIIAR, Environmental Section, P.zza Leonardo da Vinci, 32, 20133 Milano (Italy); Rigamonti, Lucia [Politecnico di Milano - DIIAR, Environmental Section, P.zza Leonardo da Vinci, 32, 20133 Milano (Italy)

    2011-09-15T23:59:59.000Z

    This paper reports the environmental results, integrated with those arising from mass and energy balances, of a research project on the comparative analysis of strategies for material and energy recovery from waste, funded by the Italian Ministry of Education, University and Research. The project, involving the cooperation of five University research groups, was devoted to the optimisation of material and energy recovery activities within integrated municipal solid waste (MSW) management systems. Four scenarios of separate collection (overall value of 35%, 50% without the collection of food waste, 50% including the collection of food waste, 65%) were defined for the implementation of energetic, environmental and economic balances. Two sizes of integrated MSW management system (IWMS) were considered: a metropolitan area, with a gross MSW production of 750,000 t/year and an average province, with a gross MSW production of 150,000 t/year. The environmental analysis was conducted using Life Cycle Assessment methodology (LCA), for both material and energy recovery activities. In order to avoid allocation we have used the technique of the expansion of the system boundaries. This means taking into consideration the impact on the environment related to the waste management activities in comparison with the avoided impacts related to the saving of raw materials and primary energy. Under the hypotheses of the study, both for the large and for the small IWMS, the energetic and environmental benefits are higher than the energetic and environmental impacts for all the scenarios analysed in terms of all the indicators considered: the scenario with 50% separate collection in a drop-off scheme excluding food waste shows the most promising perspectives, mainly arising from the highest collection (and recycling) of all the packaging materials, which is the activity giving the biggest energetic and environmental benefits. Main conclusions of the study in the general field of the assessment of the environmental performance of any integrated waste management scheme address the importance of properly defining, beyond the design value assumed for the separate collection as a whole, also the yields of each material recovered; particular significance is finally related to the amount of residues deriving from material recovery activities, resulting on average in the order of 20% of the collected materials.

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

    SciTech Connect (OSTI)

    Kapoor, A. [DOE; Gordon, S. [NSTec; Goldston, W. [Energy Solutions

    2013-07-08T23:59:59.000Z

    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 and disposal of waste.

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

    SciTech Connect (OSTI)

    Z. Ceylan

    1996-10-18T23:59:59.000Z

    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.

  19. Radium bearing waste disposal

    SciTech Connect (OSTI)

    Tope, W.G.; Nixon, D.A.; Smith, M.L.; Stone, T.J.; Vogel, R.A. [Fernald Environmental Restoration Management Corp., Cincinnati, OH (United States); Schofield, W.D. [Foster Wheeler Environmental Corp. (United States)

    1995-07-01T23:59:59.000Z

    Fernald radium bearing ore residue waste, stored within Silos 1 and 2 (K-65) and Silo 3, will be vitrified for disposal at the Nevada Test Site (NTS). A comprehensive, parametric evaluation of waste form, packaging, and transportation alternatives was completed to identify the most cost-effective approach. The impacts of waste loading, waste form, regulatory requirements, NTS waste acceptance criteria, as-low-as-reasonably-achievable principles, and material handling costs were factored into the recommended approach.

  20. Packaging and Transportation for Offsite Shipment of Materials of National Security Interest

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-12-20T23:59:59.000Z

    The purpose of this Order is to make clear that the packaging and transportation of all offsite shipments of materials of national security interest for DOE must be conducted in accordance with DOT and Nuclear Regulatory Commission (NRC) regulations that would be applicable to comparable commercial shipments, except where an alternative course of action is identified in this Order. Cancels DOE O 461.1A.

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

    SciTech Connect (OSTI)

    Marusich, Robert M.

    2013-08-15T23:59:59.000Z

    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.

  2. Processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, V.T.; Ivanov, A.V.; Filippov, E.A.

    1998-05-12T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter. 6 figs.

  3. Processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, Vitaly T. (Moscow, RU); Ivanov, Alexander V. (Moscow, RU); Filippov, Eugene A. (Moscow, RU)

    1998-05-12T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter.

  4. Method and apparatus for the management of hazardous waste material

    DOE Patents [OSTI]

    Murray, Jr., Holt (Hopewell, NJ)

    1995-01-01T23:59:59.000Z

    A container for storing hazardous waste material, particularly radioactive waste material, consists of a cylindrical body and lid of precipitation hardened C17510 beryllium-copper alloy, and a channel formed between the mated lid and body for receiving weld filler material of C17200 copper-beryllium alloy. The weld filler material has a precipitation hardening temperature lower than the aging kinetic temperature of the material of the body and lid, whereby the weld filler material is post weld heat treated for obtaining a weld having substantially the same physical, thermal, and electrical characteristics as the material of the body and lid. A mechanical seal assembly is located between an interior shoulder of the body and the bottom of the lid for providing a vacuum seal.

  5. Method and apparatus for the management of hazardous waste material

    DOE Patents [OSTI]

    Murray, H. Jr.

    1995-02-21T23:59:59.000Z

    A container for storing hazardous waste material, particularly radioactive waste material, consists of a cylindrical body and lid of precipitation hardened C17510 beryllium-copper alloy, and a channel formed between the mated lid and body for receiving weld filler material of C17200 copper-beryllium alloy. The weld filler material has a precipitation hardening temperature lower than the aging kinetic temperature of the material of the body and lid, whereby the weld filler material is post weld heat treated for obtaining a weld having substantially the same physical, thermal, and electrical characteristics as the material of the body and lid. A mechanical seal assembly is located between an interior shoulder of the body and the bottom of the lid for providing a vacuum seal. 40 figs.

  6. Definition of Small Gram Quantity Contents for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations

    SciTech Connect (OSTI)

    Sitaraman, S; Kim, S; Biswas, D; Hafner, R; Anderson, B

    2010-10-27T23:59:59.000Z

    Since the 1960's, the Department of Transportation Specification (DOT Spec) 6M packages have been used extensively for transportation of Type B quantities of radioactive materials between Department of Energy (DOE) facilities, laboratories, and productions sites. However, due to the advancement of packaging technology, the aging of the 6M packages, and variability in the quality of the packages, the DOT implemented a phased elimination of the 6M specification packages (and other DOT Spec packages) in favor of packages certified to meet federal performance requirements. DOT issued the final rule in the Federal Register on October 1, 2004 requiring that use of the DOT Specification 6M be discontinued as of October 1, 2008. A main driver for the change was the fact that the 6M specification packagings were not supported by a Safety Analysis Report for Packaging (SARP) that was compliant with Title 10 of the Code of Federal Regulations part 71 (10 CFR 71). Therefore, materials that would have historically been shipped in 6M packages are being identified as contents in Type B (and sometimes Type A fissile) package applications and addenda that are to be certified under the requirements of 10 CFR 71. The requirements in 10 CFR 71 include that the Safety Analysis Report for Packaging (SARP) must identify the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents (10 CFR 71.33(b)(1) and 10 CFR 71.33(b)(2)), and that the application (i.e., SARP submittal or SARP addendum) demonstrates that the external dose rate (due to the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents) on the surface of the packaging (i.e., package and contents) not exceed 200 mrem/hr (10 CFR 71.35(a), 10 CFR 71.47(a)). It has been proposed that a 'Small Gram Quantity' of radioactive material be defined, such that, when loaded in a transportation package, the dose rates at external points of an unshielded packaging not exceed the regulatory limits prescribed by 10 CFR 71 for non-exclusive shipments. The mass of each radioisotope presented in this paper is limited by the radiation dose rate on the external surface of the package, which per the regulatory limit should not exceed 200 mrem/hr. The results presented are a compendium of allowable masses of a variety of different isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term 'Small Gram Quantity' (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. The isotopes presented in this paper were chosen as the isotopes that Department of Energy (DOE) sites most likely need to ship. Other more rarely shipped isotopes, along with industrial and medical isotopes, are planned to be included in subsequent extensions of this work.

  7. Corrosion assessment of refractory materials for high temperature waste vitrification

    SciTech Connect (OSTI)

    Marra, J.C.; Congdon, J.W.; Kielpinski, A.L. [and others

    1995-11-01T23:59:59.000Z

    A variety of vitrification technologies are being evaluated to immobilize radioactive and hazardous wastes following years of nuclear materials production throughout the Department of Energy (DOE) complex. The compositions and physical forms of these wastes are diverse ranging from inorganic sludges to organic liquids to heterogeneous debris. Melt and off-gas products can be very corrosive at the high temperatures required to melt many of these waste streams. Ensuring material durability is required to develop viable treatment processes. Corrosion testing of materials in some of the anticipated severe environments is an important aspect of the materials identification and selection process. Corrosion coupon tests on typical materials used in Joule heated melters were completed using glass compositions with high salt contents. The presence of chloride in the melts caused the most severe attack. In the metal alloys, oxidation was the predominant corrosion mechanism, while in the tested refractory material enhanced dissolution of the refractory into the glass was observed. Corrosion testing of numerous different refractory materials was performed in a plasma vitrification system using a surrogate heterogeneous debris waste. Extensive corrosion was observed in all tested materials.

  8. Directory of Certificates of Compliance for radioactive materials packages: Report of NRC approved packages. Volume 1, Revision 18

    SciTech Connect (OSTI)

    NONE

    1995-10-01T23:59:59.000Z

    The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volumes 1 and 2. An alphabetical listing by user name is included in the back of Volume 3 of approved QA programs. The reports include a listing of all users of each package design and approved QA programs prior to the publication date.

  9. Waste minimization for commercial radioactive materials users generating low-level radioactive waste. Revision 1

    SciTech Connect (OSTI)

    Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L. [Science Applications International Corp., Idaho Falls, ID (United States)

    1991-07-01T23:59:59.000Z

    The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations. This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature.

  10. Waste minimization for commercial radioactive materials users generating low-level radioactive waste

    SciTech Connect (OSTI)

    Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L. (Science Applications International Corp., Idaho Falls, ID (United States))

    1991-07-01T23:59:59.000Z

    The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations. This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature.

  11. Material Not Categorized As Waste (MNCAW) data report. Radioactive Waste Technical Support Program

    SciTech Connect (OSTI)

    Casey, C.; Heath, B.A.

    1992-11-01T23:59:59.000Z

    The Department of Energy (DOE), Headquarters, requested all DOE sites storing valuable materials to complete a questionnaire about each material that, if discarded, could be liable to regulation. The Radioactive Waste Technical Support Program entered completed questionnaires into a database and analyzed them for quantities and type of materials stored. This report discusses the data that TSP gathered. The report also discusses problems revealed by the questionnaires and future uses of the data. Appendices contain selected data about material reported.

  12. Potential applications of nanostructured materials in nuclear waste management.

    SciTech Connect (OSTI)

    Braterman, Paul S. (The University of North Texas, Denton, TX); Phol, Phillip Isabio; Xu, Zhi-Ping (The University of North Texas, Denton, TX); Brinker, C. Jeffrey; Yang, Yi (University of New Mexico, Albuquerque, NM); Bryan, Charles R.; Yu, Kui; Xu, Huifang (University of New Mexico, Albuquerque, NM); Wang, Yifeng; Gao, Huizhen

    2003-09-01T23:59:59.000Z

    This report summarizes the results obtained from a Laboratory Directed Research & Development (LDRD) project entitled 'Investigation of Potential Applications of Self-Assembled Nanostructured Materials in Nuclear Waste Management'. The objectives of this project are to (1) provide a mechanistic understanding of the control of nanometer-scale structures on the ion sorption capability of materials and (2) develop appropriate engineering approaches to improving material properties based on such an understanding.

  13. Radioisotope thermoelectric generator package o-ring seal material validation testing

    SciTech Connect (OSTI)

    Adkins, H.E.; Ferrell, P.C.; Knight, R.C. [Westinghouse Hanford Company, P. O. Box 1970, MSIN N1-25, Richland, Washington 99352 (United States)

    1995-01-20T23:59:59.000Z

    The Radioisotope Thermoelectric Generator Package O-Ring Seal Material Validation Test was conducted to validate the use of the Butyl material as a primary seal throughout the required temperature range. Three tests were performed at (I) 233 K ({minus}40 {degree}F), (2) a specified operating temperature, and (3) 244 K ({minus}20 {degree}F) before returning to room temperature. Helium leak tests were performed at each test point to determine seal performance. The two major test objectives were to establish that butyl rubber material would maintain its integrity under various conditions and within specified parameters and to evaluate changes in material properties. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}

  14. USED NUCLEAR MATERIALS AT SAVANNAH RIVER SITE: ASSET OR WASTE?

    SciTech Connect (OSTI)

    Magoulas, V.

    2013-06-03T23:59:59.000Z

    The nuclear industry, both in the commercial and the government sectors, has generated large quantities of material that span the spectrum of usefulness, from highly valuable (“assets”) to worthless (“wastes”). In many cases, the decision parameters are clear. Transuranic waste and high level waste, for example, have no value, and is either in a final disposition path today, or – in the case of high level waste – awaiting a policy decision about final disposition. Other materials, though discardable, have intrinsic scientific or market value that may be hidden by the complexity, hazard, or cost of recovery. An informed decision process should acknowledge the asset value, or lack of value, of the complete inventory of materials, and the structure necessary to implement the range of possible options. It is important that informed decisions are made about the asset value for the variety of nuclear materials available. For example, there is a significant quantity of spent fuel available for recycle (an estimated $4 billion value in the Savannah River Site’s (SRS) L area alone); in fact, SRS has already blended down more than 300 metric tons of uranium for commercial reactor use. Over 34 metric tons of surplus plutonium is also on a path to be used as commercial fuel. There are other radiological materials that are routinely handled at the site in large quantities that should be viewed as strategically important and / or commercially viable. In some cases, these materials are irreplaceable domestically, and failure to consider their recovery could jeopardize our technological leadership or national defense. The inventories of nuclear materials at SRS that have been characterized as “waste” include isotopes of plutonium, uranium, americium, and helium. Although planning has been performed to establish the technical and regulatory bases for their discard and disposal, recovery of these materials is both economically attractive and in the national interest.

  15. Active Waste Materials Corrosion and Decontamination Tests

    SciTech Connect (OSTI)

    MJ Danielson; MR Elmore; SG Pitman

    2000-08-15T23:59:59.000Z

    Stainless steel alloys, 304L and 316L, were corrosion tested in representative radioactive samples of three actual Hanford tank waste solutions (Tanks AW-101, C-104, AN-107). Both the 304L and 316L exhibited good corrosion performance when immersed in boiling waste solutions. The maximum general corrosion rate was 0.015 mm/y (0.60 mils per year). Generally, the 304L had a slightly higher rate than the 316L. No localized attack was observed after 122 days of testing in the liquid phase, liquid/vapor phase, or vapor phase. Radioactive plate-out decontamination tests indicated that a 24-hour exposure to 1 {und M} HNO{sub 3} could remove about 99% of the radioactive components in the metal film when exposed to the C-104 and AN-107 solutions. The decontamination results are less certain for the AW-101 solution, since the initial contamination readings exceeded the capacity of the meter used for this test.

  16. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

    SciTech Connect (OSTI)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-06-01T23:59:59.000Z

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

  17. Romanian experience on packaging testing

    SciTech Connect (OSTI)

    Vieru, G. [IAEA Technical Expert, Head, Reliability and Testing Lab., Institute for Nuclear Research (Romania)

    2007-07-01T23:59:59.000Z

    With more than twenty years ago, the Institute for Nuclear Research Pitesti (INR), through its Reliability and Testing Laboratory, was licensed by the Romanian Nuclear Regulatory Body- CNCAN and to carry out qualification tests [1] for packages intended to be used for the transport and storage of radioactive materials. Radioactive materials, generated by Romanian nuclear facilities [2] are packaged in accordance with national [3] and the IAEA's Regulations [1,6] for a safe transport to the disposal center. Subjecting these packages to the normal and simulating test conditions accomplish the evaluation and certification in order to prove the package technical performances. The paper describes the qualification tests for type A and B packages used for transport and storage of radioactive materials, during a period of 20 years of experience. Testing is used to substantiate assumption in analytical models and to demonstrate package structural response. The Romanian test facilities [1,3,6] are used to simulate the required qualification tests and have been developed at INR Pitesti, the main supplier of type A packages used for transport and storage of low radioactive wastes in Romania. The testing programme will continue to be a strong option to support future package development, to perform a broad range of verification and certification tests on radioactive material packages or component sections, such as packages used for transport of radioactive sources to be used for industrial or medical purposes [2,8]. The paper describes and contain illustrations showing some of the various tests packages which have been performed during certain periods and how they relate to normal conditions and minor mishaps during transport. Quality assurance and quality controls measures taken in order to meet technical specification provided by the design there are also presented and commented. (authors)

  18. Materials evaluation programs at the Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Gee, J.T.; Iverson, D.C.; Bickford, D.F.

    1992-12-31T23:59:59.000Z

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950s to produce nuclear materials in support of the national defense effort. About 83 million gallons of high-level waste produced since operations began has been consolidated by evaporation into 33 million gallons at the waste tank farm. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF), the function of which is to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters prior to the placement of the canisters in a federal repository. The DWPF is now mechanically complete and is undergoing commissioning and run-in activities. A brief description of the DWPF process is provided.

  19. Materials evaluation programs at the Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Gee, J.T.; Iverson, D.C.; Bickford, D.F.

    1992-01-01T23:59:59.000Z

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950s to produce nuclear materials in support of the national defense effort. About 83 million gallons of high-level waste produced since operations began has been consolidated by evaporation into 33 million gallons at the waste tank farm. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF), the function of which is to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters prior to the placement of the canisters in a federal repository. The DWPF is now mechanically complete and is undergoing commissioning and run-in activities. A brief description of the DWPF process is provided.

  20. Material-not-categorized-as-waste survey for 1992

    SciTech Connect (OSTI)

    Jacobsen, P.H.

    1993-07-01T23:59:59.000Z

    In October 1992, the US Department of Energy, Richland Field Office (RL) requested that Westinghouse Hanford Company (WHC) respond to a letter from EM-331 asking for completion of a survey of items in storage but not categorized as waste (Roberts 1992). The letter contained an attachment with instructions on how to fill out the attached form and what to exclude from the survey (Appendix A). This report is a summary of the information from the response issued to RL. This report primarily is for use in estimating future waste volumes that may have been overlooked because of the nature of their classification as material not categorized as waste (MNCAW) (i.e., not yet declared Waste).

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

    SciTech Connect (OSTI)

    NONE

    1989-12-01T23:59:59.000Z

    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.

  2. Method of extruding and packaging a thin sample of reactive material, including forming the extrusion die

    DOE Patents [OSTI]

    Lewandowski, E.F.; Peterson, L.L.

    1981-11-30T23:59:59.000Z

    This invention teaches a method of cutting a narrow slot in an extrusion die with an electrical discharge machine by first drilling spaced holes at the ends of where the slot will be, whereby the oil can flow through the holes and slot to flush the material eroded away as the slot is being cut. The invention further teaches a method of extruding a very thin ribbon of solid highly reactive material such as lithium or sodium through the die in an inert atmosphere of nitrogen, argon, or the like as in a glovebox. The invention further teaches a method of stamping out sample discs from the ribbon and of packaging each disc by sandwiching it between two aluminum sheets and cold welding the sheets together along an annular seam beyond the outer periphery of the disc. This provides a sample of high purity reactive material that can have a long shelf life.

  3. Method of extruding and packaging a thin sample of reactive material including forming the extrusion die

    DOE Patents [OSTI]

    Lewandowski, Edward F. (Westmont, IL); Peterson, Leroy L. (Joliet, IL)

    1985-01-01T23:59:59.000Z

    This invention teaches a method of cutting a narrow slot in an extrusion die with an electrical discharge machine by first drilling spaced holes at the ends of where the slot will be, whereby the oil can flow through the holes and slot to flush the material eroded away as the slot is being cut. The invention further teaches a method of extruding a very thin ribbon of solid highly reactive material such as lithium or sodium through the die in an inert atmosphere of nitrogen, argon or the like as in a glovebox. The invention further teaches a method of stamping out sample discs from the ribbon and of packaging each disc by sandwiching it between two aluminum sheets and cold welding the sheets together along an annular seam beyond the outer periphery of the disc. This provides a sample of high purity reactive material that can have a long shelf life.

  4. Bentonite as a waste isolation pilot plant shaft sealing material

    SciTech Connect (OSTI)

    Daemen, J.; Ran, Chongwei [Univ. of Nevada, Reno, NV (United States)

    1996-12-01T23:59:59.000Z

    Current designs of the shaft sealing system for the Waste Isolation Pilot Plant (WIPP) propose using bentonite as a primary sealing component. The shaft sealing designs anticipate that compacted bentonite sealing components can perform through the 10,000-year regulatory period and beyond. To evaluate the acceptability of bentonite as a sealing material for the WIPP, this report identifies references that deal with the properties and characteristics of bentonite that may affect its behavior in the WIPP environment. This report reviews published studies that discuss using bentonite as sealing material for nuclear waste disposal, environmental restoration, toxic and chemical waste disposal, landfill liners, and applications in the petroleum industry. This report identifies the physical and chemical properties, stability and seal construction technologies of bentonite seals in shafts, especially in a saline brine environment. This report focuses on permeability, swelling pressure, strength, stiffness, longevity, and densification properties of bentonites.

  5. Evidence for the incorporation of lead into barite from waste rock pile materials

    E-Print Network [OSTI]

    COURTIN-NOMADE, ALEXANDRA

    2009-01-01T23:59:59.000Z

    into Barite from Waste Rock Pile Materials A L E X A N D R Awithin the waste rock pile resulting from the excavationdeveloped within waste rock pile originated from a former

  6. Injector nozzle for molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

    Brummond, W.A.; Upadhye, R.S.

    1996-02-13T23:59:59.000Z

    An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath. 2 figs.

  7. Injector nozzle for molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

    Brummond, William A. (Livermore, CA); Upadhye, Ravindra S. (Pleasanton, CA)

    1996-01-01T23:59:59.000Z

    An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath.

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

    SciTech Connect (OSTI)

    Serne, R. Jeffrey; Westsik, Joseph H.

    2011-09-05T23:59:59.000Z

    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.

  9. CH Packaging Operations Manual

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2007-05-15T23:59:59.000Z

    Introduction - This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

  10. CH Packaging Operations Manual

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2006-12-18T23:59:59.000Z

    Introduction - This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

  11. CH Packaging Operations Manual

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2007-08-22T23:59:59.000Z

    This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

  12. CH Packaging Operations Manual

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2007-11-29T23:59:59.000Z

    This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

  13. Tank waste remediation system (TWRS) privatization contractor samples waste envelope D material 241-C-106

    SciTech Connect (OSTI)

    Esch, R.A.

    1997-04-14T23:59:59.000Z

    This report represents the Final Analytical Report on Tank Waste Remediation System (TWRS) Privatization Contractor Samples for Waste Envelope D. All work was conducted in accordance with ''Addendum 1 of the Letter of Instruction (LOI) for TWRS Privatization Contractor Samples Addressing Waste Envelope D Materials - Revision 0, Revision 1, and Revision 2.'' (Jones 1996, Wiemers 1996a, Wiemers 1996b) Tank 241-C-1 06 (C-106) was selected by TWRS Privatization for the Part 1A Envelope D high-level waste demonstration. Twenty bottles of Tank C-106 material were collected by Westinghouse Hanford Company using a grab sampling technique and transferred to the 325 building for processing by the Pacific Northwest National Laboratory (PNNL). At the 325 building, the contents of the twenty bottles were combined into a single Initial Composite Material. This composite was subsampled for the laboratory-scale screening test and characterization testing, and the remainder was transferred to the 324 building for bench-scale preparation of the Privatization Contractor samples.

  14. RH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2006-11-07T23:59:59.000Z

    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 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 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. 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 conspicuous location where the activities subject to these regulations are conducted. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions or equivalent approved instructions. Following these instructions assures that operations meet the requirements of the SARP.

  15. RH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2008-01-12T23:59:59.000Z

    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 conspicuous location where the activities subject to these regulations are conducted. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions or equivalent approved instructions. Following these instructions assures that operations meet the requirements of the SARP.

  16. Estimation of residual MSW heating value as a function of waste component recycling

    SciTech Connect (OSTI)

    Magrinho, Alexandre [Mechanical Engineering Department, Escola Superior de Tecnologia de Setubal, Campus IPS, Estefanilha, Setubal (Portugal); Semiao, Viriato [Mechanical Engineering Department, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal)], E-mail: ViriatoSemiao@ist.utl.pt

    2008-12-15T23:59:59.000Z

    Recycling of packaging wastes may be compatible with incineration within integrated waste management systems. To study this, a mathematical model is presented to calculate the fraction composition of residual municipal solid waste (MSW) only as a function of the MSW fraction composition at source and recycling fractions of the different waste materials. The application of the model to the Lisbon region yielded results showing that the residual waste fraction composition depends both on the packaging wastes fraction at source and on the ratio between that fraction and the fraction of the same material, packaging and non-packaging, at source. This behaviour determines the variation of the residual waste LHV. For 100% of paper packaging recycling, LHV reduces 4.2% whereas this reduction is of 14.4% for 100% of packaging plastics recycling. For 100% of food waste recovery, LHV increases 36.8% due to the moisture fraction reduction of the residual waste. Additionally the results evidence that the negative impact of recycling paper and plastic packaging on the LHV may be compensated by recycling food waste and glass and metal packaging. This makes packaging materials recycling and food waste recovery compatible strategies with incineration within integrated waste management systems.

  17. FIFTH STATUS REPORT: TESTING OF AGED SOFTWOOD FIBERBOARD MATERIAL FOR THE 9975 SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Daugherty, W.; Skidmore, E.; Dunn, K.

    2014-04-15T23:59:59.000Z

    Samples have been prepared from a 9975 lower fiberboard subassembly fabricated from softwood fiberboard. Physical, mechanical and thermal properties have been measured following varying periods of conditioning in each of several environments. These tests have been conducted in the same manner as previous testing on cane fiberboard samples. Overall, similar aging trends are observed for softwood and cane fiberboard samples, with a few differences. Some softwood fiberboard properties tend to degrade faster in elevated humidity environments, while some cane fiberboard properties degrade faster in the hotter dry environments. As a result, it is premature to assume both materials will age at the same rates, and the preliminary aging models developed for cane fiberboard might not apply to softwood fiberboard. However, it is expected that both cane and softwood fiberboard assemblies will perform satisfactorily in conforming packages stored in a typical KAC storage environment for up to 15 years. Aging and testing of softwood fiberboard will continue and additional data will be collected. Additional samples will be added to each aging environment, to support development of an aging model specific to softwood fiberboard. Post-conditioning data have been measured on samples from a single softwood fiberboard assembly, and baseline data are also available from a limited number of vendor-provided samples. This provides minimal information on the possible sample-to-sample variation exhibited by softwood fiberboard. Data to date are generally consistent with the range seen in cane fiberboard, but some portions of the data trends are skewed toward the lower end of that range. Two additional softwood fiberboard source packages have been obtained and will begin to provide data on the range of variability of this material.

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

    SciTech Connect (OSTI)

    Reidel, Steve P.; Chamness, Mickie A.

    2007-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Colleen Shelton-Davis

    2003-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Shelton-Davis, C.V.

    2003-09-26T23:59:59.000Z

    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.

  1. Improved method and composition for immobilization of waste in cement-based material

    DOE Patents [OSTI]

    Tallent, O.K.; Dodson, K.E.; McDaniel, E.W.

    1987-10-01T23:59:59.000Z

    A composition and method for fixation or immobilization of aqueous hazardous waste material in cement-based materials (grout) is disclosed. The amount of drainable water in the cured grout is reduced by the addition of an ionic aluminum compound to either the waste material or the mixture of waste material and dry-solid cement- based material. This reduction in drainable water in the cured grout obviates the need for large, expensive amounts of gelling clays in grout materials and also results in improved consistency and properties of these cement-based waste disposal materials.

  2. DROP TESTS RESULTS OF REVISED CLOSURE BOLT CONFIGURATION OF THE STANDARD WASTE BOX, STANDARD LARGE BOX 2, AND TEN DRUM OVERPACK PACKAGINGS

    SciTech Connect (OSTI)

    May, C.; Opperman, E.; Mckeel, C.

    2010-04-15T23:59:59.000Z

    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.

  3. Construction materials as a waste management solution for cellulose sludge

    SciTech Connect (OSTI)

    Modolo, R., E-mail: regina.modolo@ua.pt [University of Aveiro, Civil Engineering Department/CICECO, 3810-193 Aveiro (Portugal); Ferreira, V.M. [University of Aveiro, Civil Engineering Department/CICECO, 3810-193 Aveiro (Portugal); Machado, L.M. [RAIZ - Forest and Paper Research Institute, Portucel-Soporcel, Eixo (Portugal); Rodrigues, M.; Coelho, I. [CIMIANTO - Sociedade Tecnica Hidraulica, S.A., Alhandra (Portugal)

    2011-02-15T23:59:59.000Z

    Sustainable waste management system for effluents treatment sludge has been a pressing issue for pulp and paper sector. Recycling is always recommended in terms of environmental sustainability. Following an approach of waste valorisation, this work aims to demonstrate the technical viability of producing fiber-cement roof sheets incorporating cellulose primary sludge generated on paper and pulp mills. From the results obtained with preliminary studies it was possible to verify the possibility of producing fiber-cement sheets by replacing 25% of the conventional used virgin long fiber by primary effluent treatment cellulose sludge. This amount of incorporation was tested on an industrial scale. Environmental parameters related to water and waste, as well as tests for checking the quality of the final product was performed. These control parameters involved total solids in suspension, dissolved salts, chlorides, sulphates, COD, metals content. In the product, parameters like moisture, density and strength were controlled. The results showed that it is possible to replace the virgin long fibers pulp by primary sludge without impacts in final product characteristics and on the environment. This work ensures the elimination of significant waste amounts, which are nowadays sent to landfill, as well as reduces costs associated with the standard raw materials use in the fiber-cement industrial sector.

  4. Procedure for the Recycling Material and Disposal of Waste from

    E-Print Network [OSTI]

    Guillas, Serge

    assessments must include consideration of storage, handling, movement and disposal of wastes under that waste is produced, stored, transported and disposed of without harming the environment. This is your Clinical Wastes Radioactive Wastes Laboratory Wastes of Unknown Hazard Non-Hazardous Laboratory Wastes

  5. PATRAM '92: 10th international symposium on the packaging and transportation of radioactive materials [Papers presented by Sandia National Laboratories

    SciTech Connect (OSTI)

    none,

    1992-01-01T23:59:59.000Z

    This document provides the papers presented by Sandia Laboratories at PATRAM '92, the tenth International symposium on the Packaging and Transportation of Radioactive Materials held September 13--18, 1992 in Yokohama City, Japan. Individual papers have been cataloged separately. (FL)

  6. Northwest Plume Groundwater System Green-sand Media Removal and Waste Packaging Paducah Gaseous Diffusion Plant, Paducah, Kentucky

    SciTech Connect (OSTI)

    Troutman, M.T.; Richards, C.J.; Tarantino, J.J. [CDM Federal Programs Corporation, 325 Kentucky Avenue, Kevil, KY 42053 (United States)

    2006-07-01T23:59:59.000Z

    The Northwest Plume Groundwater System (NWPGS) was temporarily shut down due to high differential pressures across the green-sand filters. Increased levels of suspended solids were introduced into the system from monitoring well development water, equipment decontamination water, and secondary containment water. These waters were treated for suspended solids through a groundwater pretreatment system but were suspected of causing the high differential pressures in the green-sand filters. Prior to the system being shutdown, the NWPGS had been experiencing increasingly shorter run times between filter backwashes indicating that the normal backwash cycle was not adequately removing the fines. This condition led to the removal and replacement of green-sand media from two filter vessels. Discussions include problems with the removal process, waste packaging specifications, requirements for the disposition of green-sand media, and lessons learned. (authors)

  7. Method for co-processing waste rubber and carbonaceous material

    DOE Patents [OSTI]

    Farcasiu, Malvina (Pittsburgh, PA); Smith, Charlene M. (Pittsburgh, PA)

    1991-01-01T23:59:59.000Z

    In a process for the co-processing of waste rubber and carbonaceous material to form a useful liquid product, the rubber and the carbonaceous material are combined and heated to the depolymerization temperature of the rubber in the presence of a source of hydrogen. The depolymerized rubber acts as a liquefying solvent for the carbonaceous material while a beneficial catalytic effect is obtained from the carbon black released on depolymerization the reinforced rubber. The reaction is carried out at liquefaction conditions of 380.degree.-600.degree. C. and 70-280 atmospheres hydrogen pressure. The resulting liquid is separated from residual solids and further processed such as by distillation or solvent extraction to provide a carbonaceous liquid useful for fuels and other purposes.

  8. Method for co-processing waste rubber and carbonaceous material

    SciTech Connect (OSTI)

    Farcasiu, M.; Smith, C.M.

    1990-10-09T23:59:59.000Z

    In a process for the co-processing of waste rubber and carbonaceous material to form a useful liquid product, the rubber and the carbonaceous material are combined and heated to the depolymerization temperature of the rubber in the presence of a source of hydrogen. The deploymerized rubber acts as a liquefying solvent for the carbonaceous material while a beneficial catalytic effect is obtained from the carbon black released on deploymerization the reinforced rubber. The reaction is carried out at liquefaction conditions of 380--600{degrees}C and 70--280 atmospheres hydrogen pressure. The resulting liquid is separated from residual solids and further processed such as by distillation or solvent extraction to provide a carbonaceous liquid useful for fuels and other purposes.

  9. Application of United States Department of Transportation regulations to hazardous material and waste shipments on the Hanford Site

    SciTech Connect (OSTI)

    Burnside, M.E.

    1992-01-01T23:59:59.000Z

    All hazardous material and waste transported over roadways open to the public must be in compliance with the US Department of Transportation (DOT) regulations. The DOT states that the hazardous material regulations (HMR) also apply to government-owned, contractor-operated (GOCO) transportation operations over any US Department of Energy (DOE) site roadway where the public has free and unrestricted access. Hazardous material and waste in packages that do not meet DOE regulations must be transported on DOE site roadways in a manner that excludes the public and nonessential workers. At the DOE Richland Field Office (the Hanford Site), hazardous material and waste movements that do not meet DOE requirements are transported over public access roadways during off-peak hours with the roadways barricaded. These movements are accomplished using a transportation plan that involves the DOE, DOE contractors, and private utilities who operate on or near the Hanford Site. This method, which is used at the Hanford Site to comply with DOE regulations onsite, can be communicated to other DOE sites to provide a basis for achieving consistency in similar transportation operations.

  10. Development of New Low-Cost, High-Performance, PV Module Encapsulant/Packaging Materials: Final Technical Progress Report, 22 October 2002 - 15 November 2007

    SciTech Connect (OSTI)

    Tucker, R.

    2008-04-01T23:59:59.000Z

    Report on objectives to work with U.S.-based PV module manufacturers (c-Si, a-Si, CIS, other thin films) to develop/qualify new low-cost, high-performance PV module encapsulant/packaging materials, and processes using the packaging materials.

  11. Characterisation of Plasma Vitrified Simulant Plutonium Contaminated Material Waste

    SciTech Connect (OSTI)

    Hyatt, Neil C.; Morgan, Suzy; Stennett, Martin C. [Immobilisation Science Laboratory, Department of Engineering Materials, The University of Sheffield, Mappin Street, Sheffield, S1 3JD (United Kingdom); Scales, Charlie R. [Nexia Solutions Ltd., Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom); Deegan, David [Tetronics Ltd., 5, Lechlade Road, Faringdon, Oxfordshire, SN7 8AL (United Kingdom)

    2007-07-01T23:59:59.000Z

    The potential of plasma vitrification for the treatment of a simulant Plutonium Contaminated Material (PCM) was investigated. It was demonstrated that the PuO{sub 2} simulant, CeO{sub 2}, could be vitrified in the amorphous calcium iron aluminosilicate component of the product slag with simultaneous destruction of the organic and polymer waste fractions. Product Consistency Tests conducted at 90 deg. C in de-ionised water and buffered pH 11 solution show the PCM slag product to be durable with respect to release of Ce. (authors)

  12. Electronic waste (e-waste): Material flows and management practices in Nigeria

    SciTech Connect (OSTI)

    Nnorom, Innocent Chidi [Department of Industrial Chemistry, Abia State University, Uturu, Abia State (Nigeria)], E-mail: chidiabsu@yahoo.co.uk; Osibanjo, Oladele [Basel Convention Regional Coordinating Center for Africa for Training and Technology Transfer, Department of Chemistry, University of Ibadan (Nigeria)], E-mail: osibanjo@baselnigeria.com

    2008-07-01T23:59:59.000Z

    The growth in electrical and electronic equipment (EEE) production and consumption has been exponential in the last two decades. This has been as a result of the rapid changes in equipment features and capabilities, decrease in prices, and the growth in internet use. This creates a large volume of waste stream of obsolete electrical and electronic devices (e-waste) in developed countries. There is high level of trans-boundary movement of these devices as secondhand electronic equipment into developing countries in an attempt to bridge the 'digital divide'. The past decade has witnessed a phenomenal advancement in information and communication technology (ICT) in Nigeria, most of which rely on imported secondhand devices. This paper attempts to review the material flow of secondhand/scrap electronic devices into Nigeria, the current management practices for e-waste and the environmental and health implications of such low-end management practices. Establishment of formal recycling facilities, introduction of legislation dealing specifically with e-waste and the confirmation of the functionality of secondhand EEE prior to importation are some of the options available to the government in dealing with this difficult issue.

  13. DOE Waste Package Project. Quarterly progress report, January 1, 1995--March 31, 1995

    SciTech Connect (OSTI)

    Ladkany, S.G.

    1995-05-01T23:59:59.000Z

    Research progress is reported on the design of containers for high-level radioactive wastes to be emplaced at the Yucca Mountain underground repository. Tasks included: temperature distribution and heat flow around the containers; failure possibility due to mechanical stresses and pitting corrosion; robotic manipulation of the containers; and design requirements of rock tunnel drift for long term storage.

  14. MIXING OF INCOMPATIBLE MATERIALS IN WASTE TANKS TECHNICAL BASIS DOCUMENT

    SciTech Connect (OSTI)

    SANDGREN, K.R.

    2003-10-15T23:59:59.000Z

    This document presents onsite radiological, onsite toxicological, and offsite toxicological consequences, risk binning, and control decision results for the mixing of incompatible materials in waste tanks representative accident. This technical basis document was developed to support the tank farms documented safety analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls.

  15. Benefits/impacts of utilizing depleted uranium silicate glass as backfill for spent fuel waste packages

    SciTech Connect (OSTI)

    Pope, R.B.; Forsberg, C.W.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1996-05-01T23:59:59.000Z

    An assessment has been made of the benefits and impacts which can be derived by filling a spent nuclear fuel multi-purpose canister with depleted uranium silicate (DUS) glass at a reactor site. Although the primary purpose of the DUS glass fill would be to enhance repository performance assessment and control criticality of geologic times, a number of benefits to the waste management system can be derived from adding the DUS glass prior to shipment from the reactor site.

  16. Directory of Certificates of Compliance for radioactive materials packages: Certificates of Compliance. Volume 2, Revision 18

    SciTech Connect (OSTI)

    NONE

    1995-10-01T23:59:59.000Z

    The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volumes 1 and 2. An alphabetical listing by user name is included in the back of Volume 3 of approved QA programs. The reports include a listing of all users of each package design and approved QA programs prior to the publication date.

  17. The tenth conference on solid waste management & materials policy and the New York State solid waste management

    SciTech Connect (OSTI)

    NONE

    1995-05-01T23:59:59.000Z

    The proceedings of the Tenth Conference on Solid Waste Management and Materials Policy and the New York State Solid Waste Management held February 19-22, 1995 in New York City are presented. Such topics as recycling, resource recovery, emission characteristics of burn barrels, ash management, controlling landfill closure costs, flow control and federalism, composting programs, air pollutant emissions from MSW landfills, backyard waste management, waste-based manufacturing, and scrap tire management are covered. A separate abstract and indexing were prepared for each paper for inclusion in the Energy Science and Technology Database.

  18. Sepiolite as an Alternative Liner Material in Municipal Solid Waste Landfills

    E-Print Network [OSTI]

    Aydilek, Ahmet

    Sepiolite as an Alternative Liner Material in Municipal Solid Waste Landfills Yucel Guney1 ; Savas in municipal solid waste landfills. However, natural clays may not always provide good contaminant sorption in solid waste landfills. DOI: 10.1061/ ASCE 1090-0241 2008 134:8 1166 CE Database subject headings

  19. Recharge Data Package for Hanford Single-Shell Tank Waste Management Areas

    SciTech Connect (OSTI)

    Fayer, Michael J.; Keller, Jason M.

    2007-09-24T23:59:59.000Z

    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.

  20. 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 (OSTI)

    J. M. Capron

    2008-06-16T23:59:59.000Z

    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.

  1. Safety analysis report for packaging (Onsite) for the Hanford Ecorok packaging

    SciTech Connect (OSTI)

    Mercado, M.S.

    1996-02-23T23:59:59.000Z

    This safety analysis report for packaging approves the Hanford Ecorok packaging for shipping contaminated water purification filters from K Basins to the Central Waste Complex.

  2. Method for acid oxidation of radioactive, hazardous, and mixed organic waste materials

    DOE Patents [OSTI]

    Pierce, Robert A. (Aiken, SC); Smith, James R. (Corrales, NM); Ramsey, William G. (Aiken, SC); Cicero-Herman, Connie A. (Aiken, SC); Bickford, Dennis F. (Folly Beach, SC)

    1999-01-01T23:59:59.000Z

    The present invention is directed to a process for reducing the volume of low level radioactive and mixed waste to enable the waste to be more economically stored in a suitable repository, and for placing the waste into a form suitable for permanent disposal. The invention involves a process for preparing radioactive, hazardous, or mixed waste for storage by contacting the waste starting material containing at least one organic carbon-containing compound and at least one radioactive or hazardous waste component with nitric acid and phosphoric acid simultaneously at a contacting temperature in the range of about 140.degree. C. to about 210 .degree. C. for a period of time sufficient to oxidize at least a portion of the organic carbon-containing compound to gaseous products, thereby producing a residual concentrated waste product containing substantially all of said radioactive or inorganic hazardous waste component; and immobilizing the residual concentrated waste product in a solid phosphate-based ceramic or glass form.

  3. 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 (OSTI)

    Gonzales D.

    2010-03-30T23:59:59.000Z

    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.

  4. Waste segregation procedures and benefits

    SciTech Connect (OSTI)

    Fish, J.D.; Massey, C.D.; Ward, S.J.

    1990-01-01T23:59:59.000Z

    Segregation is a critical first step in handling hazardous and radioactive materials to minimize the generation of regulated wastes. In addition, segregation can significantly reduce the complexity and the total cost of managing waste. Procedures at Sandia National Laboratories, Albuquerque require that wastes be segregated, first, by waste type (acids, solvents, low level radioactive, mixed, classified, etc.). Higher level segregation requirements, currently under development, are aimed at enhancing the possibilities for recovery, recycle and reapplication; reducing waste volumes; reducing waste disposal costs, and facilitating packaging storage, shipping and disposal. 2 tabs.

  5. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1997-03-01T23:59:59.000Z

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs.

  6. Waste Receiving and Packaging, Module 2A, Supplemental Design Requirements Document

    SciTech Connect (OSTI)

    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.

    1994-04-26T23:59:59.000Z

    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.

  7. Permit Fees for Hazardous Waste Material Management (Connecticut)

    Broader source: Energy.gov [DOE]

    These regulations describe applicable fees for permit application, modification, and transfer for permits related to hazardous waste management.

  8. Accepting Mixed Waste as Alternate Feed Material for Processing and Disposal at a Licensed Uranium Mill

    SciTech Connect (OSTI)

    Frydenland, D. C.; Hochstein, R. F.; Thompson, A. J.

    2002-02-26T23:59:59.000Z

    Certain categories of mixed wastes that contain recoverable amounts of natural uranium can be processed for the recovery of valuable uranium, alone or together with other metals, at licensed uranium mills, and the resulting tailings permanently disposed of as 11e.(2) byproduct material in the mill's tailings impoundment, as an alternative to treatment and/or direct disposal at a mixed waste disposal facility. This paper discusses the regulatory background applicable to hazardous wastes, mixed wastes and uranium mills and, in particular, NRC's Alternate Feed Guidance under which alternate feed materials that contain certain types of mixed wastes may be processed and disposed of at uranium mills. The paper discusses the way in which the Alternate Feed Guidance has been interpreted in the past with respect to processing mixed wastes and the significance of recent changes in NRC's interpretation of the Alternate Feed Guidance that sets the stage for a broader range of mixed waste materials to be processed as alternate feed materials. The paper also reviews the le gal rationale and policy reasons why materials that would otherwise have to be treated and/or disposed of as mixed waste, at a mixed waste disposal facility, are exempt from RCRA when reprocessed as alternate feed material at a uranium mill and become subject to the sole jurisdiction of NRC, and some of the reasons why processing mixed wastes as alternate feed materials at uranium mills is preferable to direct disposal. Finally, the paper concludes with a discussion of the specific acceptance, characterization and certification requirements applicable to alternate feed materials and mixed wastes at International Uranium (USA) Corporation's White Mesa Mill, which has been the most active uranium mill in the processing of alternate feed materials under the Alternate Feed Guidance.

  9. In-plant recycling of ironmaking waste materials at Pohang Works

    SciTech Connect (OSTI)

    Kim, C.H.; Jung, S. [POSCO, Pohang (Korea, Republic of). Ironmaking Dept.

    1997-12-31T23:59:59.000Z

    The regulations for pollution control are being strengthened more year by year. Therefore, waste materials containing iron oxides are being increasingly used in the sinter plant. As a result, waste materials recycling in the sintering process not only reduces costs by eliminating waste disposal costs and utilizing Fe bearing by-products to replace iron ores and flux materials, but gives fuel rate benefits to the sintering process through heat of oxidizing of Fe bearing materials and combustion of coke fines carried with Fe Bearing by-products.

  10. Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials

    SciTech Connect (OSTI)

    Lindle, Dennis W.

    2011-04-21T23:59:59.000Z

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization. Specially formulated glass compositions and ceramics such as pyrochlores and apatites are the main candidates for these wastes. An important consideration linked to the durability of waste-form materials is the local structure around the waste components. Equally important is the local structure of constituents of the glass and ceramic host matrix. Knowledge of the structure in the waste-form host matrices is essential, prior to and subsequent to waste incorporation, to evaluate and develop improved waste-form compositions based on scientific considerations. This project used the soft-x-ray synchrotron-radiation-based technique of near-edge x-ray-absorption fine structure (NEXAFS) as a unique method for investigating oxidation states and structures of low-Z elemental constituents forming the backbones of glass and ceramic host matrices for waste-form materials. In addition, light metal ions in ceramic hosts, such as titanium, are also ideal for investigation by NEXAFS in the soft-x-ray region. Thus, one of the main objectives was to understand outstanding issues in waste-form science via NEXAFS investigations and to translate this understanding into better waste-form materials, followed by eventual capability to investigate “real” waste-form materials by the same methodology. We conducted several detailed structural investigations of both pyrochlore ceramic and borosilicate-glass materials during the project and developed improved capabilities at Beamline 6.3.1 of the Advanced Light Source (ALS) to perform the studies.

  11. Qualitative and Quantitative Assessment of Nuclear Materials Contained in High-Activity Waste Arising from the Operations at the 'SHELTER' Facility

    SciTech Connect (OSTI)

    Cherkas, Dmytro

    2011-10-01T23:59:59.000Z

    As a result of the nuclear accident at the Chernobyl NPP in 1986, the explosion dispeesed nuclear materials contained in the nuclear fuel of the reactor core over the destroyed facilities at Unit No. 4 and over the territory immediately adjacent to the destroyed unit. The debris was buried under the Cascade Wall. Nuclear materials at the SHELTER can be characterized as spent nuclear fuel, fresh fuel assemblies (including fuel assemblies with damaged geometry and integrity, and individual fuel elements), core fragments of the Chernobyl NPP Unit No. 4, finely-dispersed fuel (powder/dust), uranium and plutonium compounds in water solutions, and lava-like nuclear fuel-containing masses. The new safe confinement (NSC) is a facility designed to enclose the Chernobyl NPP Unit No. 4 destroyed by the accident. Construction of the NSC involves excavating operations, which are continuously monitored including for the level of radiation. The findings of such monitoring at the SHELTER site will allow us to characterize the recovered radioactive waste. When a process material categorized as high activity waste (HAW) is detected the following HLW management operations should be involved: HLW collection; HLW fragmentation (if appropriate); loading HAW into the primary package KT-0.2; loading the primary package filled with HAW into the transportation cask KTZV-0.2; and storing the cask in temporary storage facilities for high-level solid waste. The CDAS system is a system of 3He tubes for neutron coincidence counting, and is designed to measure the percentage ratio of specific nuclear materials in a 200-liter drum containing nuclear material intermixed with a matrix. The CDAS consists of panels with helium counter tubes and a polyethylene moderator. The panels are configured to allow one to position a waste-containing drum and a drum manipulator. The system operates on the ‘add a source’ basis using a small Cf-252 source to identify irregularities in the matrix during an assay. The platform with the source is placed under the measurement chamber. The platform with the source material is moved under the measurement chamber. The design allows one to move the platform with the source in and out, thus moving the drum. The CDAS system and radioactive waste containers have been built. For each drum filled with waste two individual measurements (passive/active) will be made. This paper briefly describes the work carried out to assess qualitatively and quantitatively the nuclear materials contained in high-level waste at the SHELTER facility. These efforts substantially increased nuclear safety and security at the facility.

  12. RADIOACTIVE MATERIAL SHIPPING PACKAGINGS AND METAL TO METAL SEALS FOUND IN THE CLOSURES OF CONTAINMENT VESSELS INCORPORATING CONE SEAL CLOSURES

    SciTech Connect (OSTI)

    Loftin, B; Glenn Abramczyk, G; Allen Smith, A

    2007-06-06T23:59:59.000Z

    The containment vessels for the Model 9975 radioactive material shipping packaging employ a cone-seal closure. The possibility of a metal-to-metal seal forming between the mating conical surfaces, independent of the elastomer seals, has been raised. It was postulated that such an occurrence would compromise the containment vessel hydrostatic and leakage tests. The possibility of formation of such a seal has been investigated by testing and by structural and statistical analyses. The results of the testing and the statistical analysis demonstrate and procedural changes ensure that hydrostatic proof and annual leakage testing can be accomplished to the appropriate standards.

  13. Data summary of municipal solid waste management alternatives. Volume 7, Appendix E -- Material recovery/material recycling technologies

    SciTech Connect (OSTI)

    none,

    1992-10-01T23:59:59.000Z

    The enthusiasm for and commitment to recycling of municipal solid wastes is based on several intuitive benefits: Conservation of landfill capacity; Conservation of non-renewable natural resources and energy sources; Minimization of the perceived potential environmental impacts of MSW combustion and landfilling; Minimization of disposal costs, both directly and through material resale credits. In this discussion, ``recycling`` refers to materials recovered from the waste stream. It excludes scrap materials that are recovered and reused during industrial manufacturing processes and prompt industrial scrap. Materials recycling is an integral part of several solid waste management options. For example, in the preparation of refuse-derived fuel (RDF), ferrous metals are typically removed from the waste stream both before and after shredding. Similarly, composting facilities, often include processes for recovering inert recyclable materials such as ferrous and nonferrous metals, glass, Plastics, and paper. While these two technologies have as their primary objectives the production of RDF and compost, respectively, the demonstrated recovery of recyclables emphasizes the inherent compatibility of recycling with these MSW management strategies. This appendix discusses several technology options with regard to separating recyclables at the source of generation, the methods available for collecting and transporting these materials to a MRF, the market requirements for post-consumer recycled materials, and the process unit operations. Mixed waste MRFs associated with mass bum plants are also presented.

  14. Process and material that encapsulates solid hazardous waste

    DOE Patents [OSTI]

    O'Brien, Michael H. (Idaho Falls, ID); Erickson, Arnold W. (Idaho Falls, ID)

    1999-01-01T23:59:59.000Z

    A method of encapsulating mixed waste in which a thermoplastic polymer having a melting temperature less than about 150.degree. C. and sulfur and mixed waste are mixed at an elevated temperature not greater than about 200.degree. C. and mixed for a time sufficient to intimately mix the constituents, and then cooled to a solid. The resulting solid is also disclosed.

  15. Packaging Materials of the 21st Century: "Sustainable Nano-Materials - Benefits to the industry"

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin of Contamination in235-1 Termoelectrica U.SPRESS FACTBiofuels LLCAvailable at thePackaging

  16. MATERIAL FLUX ANALYSIS (MFA) FOR PLANNING OF DOMESTIC WASTES AND WASTEWATER MANAGEMENT

    E-Print Network [OSTI]

    Richner, Heinz

    i MATERIAL FLUX ANALYSIS (MFA) FOR PLANNING OF DOMESTIC WASTES AND WASTEWATER MANAGEMENT: CASE of Nonthaburi, Statistical office of Nonthaburi and Agricultural extension office of Pak Kret for their kind nutrient management, organic waste, wastewater and septage that contained high concentration of nutrients

  17. Cleanup Verification Package for the 100-F-20, Pacific Northwest Laboratory Parallel Pits

    SciTech Connect (OSTI)

    M. J. Appel

    2007-01-22T23:59:59.000Z

    This cleanup verification package documents completion of remedial action for the 100-F-20, Pacific Northwest Laboratory Parallel Pits waste site. This waste site consisted of two earthen trenches thought to have received both radioactive and nonradioactive material related to the 100-F Experimental Animal Farm.

  18. Reverse logistics and large-scale material recovery from electronics waste

    E-Print Network [OSTI]

    Krones, Jonathan Seth

    2007-01-01T23:59:59.000Z

    Waste consolidation is a crucial step in the development of cost-effective, nation-wide material reclamation networks. This thesis project investigates typical and conformational tendencies of a hypothetical end-of-life ...

  19. Safety Requirements for the Packaging and Transportation of Hazardous Materials, Hazardous Substances, and Hazardous Wastes

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1985-07-09T23:59:59.000Z

    Cancels Chapter 3 of DOE 5480.1A. Canceled by DOE O 460.1 of 9-27-1995 and by DOE N 251.4 & Para. 9c canceled by DOE O 231.1 of 9-30-1995.

  20. Materials and Security Consolidation Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect (OSTI)

    Not Listed

    2011-09-01T23:59:59.000Z

    Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Security Consolidation Center facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

  1. Materials and Fuels Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect (OSTI)

    Lisa Harvego; Brion Bennett

    2011-09-01T23:59:59.000Z

    Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Fuels Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

  2. Packaging and Transfer or Transportation of Materials of National Security Interest

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2004-04-26T23:59:59.000Z

    To establish requirements and responsibilities for offsite shipments of naval nuclear fuel elements, Category I and Category II special nuclear material, nuclear explosives, nuclear components, special assemblies, and other materials of national security interest. Cancels DOE O 461.1. Canceled by DOE O 461.1B and DOE O 461.2.

  3. Preliminary waste acceptance criteria for the ICPP spent fuel and waste management technology development program

    SciTech Connect (OSTI)

    Taylor, L.L.; Shikashio, R.

    1993-09-01T23:59:59.000Z

    The purpose of this document is to identify requirements to be met by the Producer/Shipper of Spent Nuclear Fuel/High-LeveL Waste SNF/HLW in order for DOE to be able to accept the packaged materials. This includes defining both standard and nonstandard waste forms.

  4. Performance analysis of co-firing waste materials in an advanced pressurized fluidized-bed combustor

    SciTech Connect (OSTI)

    Bonk, D.L.; McDaniel, H.M. [USDOE Morgantown Energy Technology Center, WV (United States); DeLallo, M.R. Jr.; Zaharchuk, R. [Gilbert/Commonwealth, Inc., Reading, PA (United States)

    1995-07-01T23:59:59.000Z

    The co-firing of waste materials with coal in utility scale power plants has emerged as an effective approach to produce energy and manage municipal wastes. Leading this approach is the atmospheric fluidized-bed combustor (AFBC). It has demonstrated its commercial acceptance in the utility market as a reliable source of power by burning a variety of waste and alternative fuels. The application of pressurized fluidized-bed combustor (PFBC) technology, although relatively new, can provide significant enhancements to the efficient production of electricity while maintaining the waste management benefits of AFBC. A study was undertaken to investigate the technical and economical feasibility of co-firing a PFBC with coal and municipal and industrial wastes. Focus was placed on the production of electricity and the efficient disposal of wastes for application in central power station and distributed locations. Issues concerning waste material preparation and feed, PFBC operation, plant emissions, and regulations are addressed. The results and conclusions developed are generally applicable to current and advanced PFBC design concepts. Wastes considered for co-firing include municipal solid waste (MSW), sewage sludge, and industrial de-inking sludge. Conceptual designs of two power plants rated at 250 MWe and 150 MWe were developed. Heat and material balances were completed for each plant along with environmental issues. With the PFBC`s operation at high temperature and pressure, efforts were centered on defining feeding systems capable of operating at these conditions. Air emissions and solid wastes were characterized to assess the environmental performance comparing them to state and Federal regulations. This paper describes the results of this investigation, presents conclusions on the key issues, and provides recommendations for further evaluation.

  5. The critical role of microscopy and spectroscopy in the development of new materials for microelectronics packaging

    SciTech Connect (OSTI)

    Youngman, R.A. [Carborundum Corp., Phoenix, AZ (United States)

    1996-12-31T23:59:59.000Z

    It has been over thirty years since sintered aluminum nitride (AlN) has been the focus of many research and development activities in Japan, the U.S., and Europe. Only in the past 5 years has there been significant use of this material in microelectronics. There are many reasons for this considerable time for application including, technology needs and acceptance of a new material. Also important has been the role of materials understanding of AlN through the use of microscopy and spectroscopy. We illustrate the use of both standard and unique characterization techniques to elucidate the nature of the crystalline defects which control the important property of thermal conductivity.

  6. Ion-exchange material and method of storing radioactive wastes

    DOE Patents [OSTI]

    Komarneni, S.; Roy, D.M.

    1983-10-31T23:59:59.000Z

    A new cation exchanger is a modified tobermorite containing aluminum isomorphously substituted for silicon and containing sodium or potassium. The exchanger is selective for lead, rubidium, cobalt, and cadmium and is selective for cesium over calcium or sodium. The tobermorites are compatible with cement and are useful for the long-term fixation and storage of radioactive nuclear wastes.

  7. Apparatus for the processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, V.T.; Ivanov, A.V.; Filippov, E.A.

    1999-03-16T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter. 6 figs.

  8. Apparatus for the processing of solid mixed waste containing radioactive and hazardous materials

    DOE Patents [OSTI]

    Gotovchikov, Vitaly T. (Moscow, RU); Ivanov, Alexander V. (Moscow, RU); Filippov, Eugene A. (Moscow, RU)

    1999-03-16T23:59:59.000Z

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination oaf plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter.

  9. Utilizing New Binder Materials for Green Building has Zero Waste by Recycling Slag and Sewage Sludge Ash 

    E-Print Network [OSTI]

    Zeedan, S. R.

    2010-01-01T23:59:59.000Z

    binding material to save energy and to produce new innovative zero materials waste . The current research aims to investigate new binder materials as alternative of Portland cement. Alkali activated slag (AAS) blended with sewage sludge ash (SSA...

  10. Materials selection for process equipment in the Hanford waste vitrification plant

    SciTech Connect (OSTI)

    Elmore, M R; Jensen, G A

    1991-07-01T23:59:59.000Z

    The Hanford Waste Vitrification Plant (HWVP) is being designed to vitrify defense liquid high-level wastes and transuranic wastes stored at Hanford. The HWVP Functional Design Criteria (FDC) requires that materials used for fabrication of remote process equipment and piping in the facility be compatible with the expected waste stream compositions and process conditions. To satisfy FDC requirements, corrosion-resistant materials have been evaluated under simulated HWVP-specific conditions and recommendations have been made for HWVP applications. The materials recommendations provide to the project architect/engineer the best available corrosion rate information for the materials under the expected HWVP process conditions. Existing data and sound engineering judgement must be used and a solid technical basis must be developed to define an approach to selecting suitable construction materials for the HWVP. This report contains the strategy, approach, criteria, and technical basis developed for selecting materials of construction. Based on materials testing specific to HWVP and on related outside testing, this report recommends for constructing specific process equipment and identifies future testing needs to complete verification of the performance of the selected materials. 30 refs., 7 figs., 11 tabs.

  11. Advanced Thermoelectric Materials for Efficient Waste Heat Recovery in Process Industries

    SciTech Connect (OSTI)

    Adam Polcyn; Moe Khaleel

    2009-01-06T23:59:59.000Z

    The overall objective of the project was to integrate advanced thermoelectric materials into a power generation device that could convert waste heat from an industrial process to electricity with an efficiency approaching 20%. Advanced thermoelectric materials were developed with figure-of-merit ZT of 1.5 at 275 degrees C. These materials were not successfully integrated into a power generation device. However, waste heat recovery was demonstrated from an industrial process (the combustion exhaust gas stream of an oxyfuel-fired flat glass melting furnace) using a commercially available (5% efficiency) thermoelectric generator coupled to a heat pipe. It was concluded that significant improvements both in thermoelectric material figure-of-merit and in cost-effective methods for capturing heat would be required to make thermoelectric waste heat recovery viable for widespread industrial application.

  12. Evaluation of Basic Parameters for Packaging, Storage and Transportation of Biomass Material from Field to Biorefinery

    E-Print Network [OSTI]

    Paliwal, Richa

    2012-02-14T23:59:59.000Z

    Logistics and Building a Function Structure for Biomass Materials????????????????????.. 14 2.6.1 Cotton Logistics ....................................................................... 14 2... tolerance, and its adaptability to a wide range of soils have made it 6 a potential energy crop [2, 4, 14, 15]. Cundiff et al. [16] also states the benefit of using switchgrass: it can be harvested and baled using conventional agricultural equipment...

  13. A Strategy for Quantifying Radioactive Material in a Low-Level Waste Incineration Facility

    SciTech Connect (OSTI)

    Hochel, R.C. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1997-03-01T23:59:59.000Z

    One of the methods proposed by the U.S. Department of Energy (DOE) for the volume reduction and stabilization of a variety of low-level radioactive wastes (LLW) is incineration. Many commercial incinerators are in operation treating both non-hazardous and hazardous wastes. These can obtain volume reductions factors of 50 or more for certain wastes, and produce a waste (ash) that can be easily stabilized if necessary by vitrification or cementation. However, there are few incinerators designed to accommodate radioactive wastes. One has been recently built at the Savannah River Site (SRS) near Aiken, SC and is burning non-radioactive hazardous waste and radioactive wastes in successive campaigns. The SRS Consolidated Incineration Facility (CIF) is RCRA permitted as a Low Chemical Hazard, Radiological facility as defined by DOE criteria (Ref. 1). Accordingly, the CIF must operate within specified chemical, radionuclide, and fissile material inventory limits (Ref. 2). The radionuclide and fissile material limits are unique to radiological or nuclear facilities, and require special measurement and removal strategies to assure compliance, and the CIF may be required to shut down periodically in order to clean out the radionuclide inventory which builds up in various parts of the facility.

  14. Economics of co-firing waste materials in an advanced pressurized fluidized-bed combustor

    SciTech Connect (OSTI)

    Bonk, D.L.; McDaniel, H.M. [Dept. of Energy, Morgantown, WV (United States). Morgantown Energy Technology Center; DeLallo, M.R. Jr.; Zaharchuk, R. [Gilbert/Commonwealth, Inc., Reading, PA (United States)

    1995-04-01T23:59:59.000Z

    The co-firing of waste materials with coal in utility scale power plants has emerged as an effective approach to produce energy and manage municipal waste. Leading this approach is the atmospheric fluidized bed combustor (AFBC). It has demonstrated its commercial acceptance in the utility market as a reliable source of power by burning a variety of waste and alternative fuels. The fluidized bed, with its stability of combustion, reduces the amount of thermochemical transients and provides for easier process control. The application of pressurized fluidized-bed combustor (PFBC) technology, although relatively new, can provide significant enhancements to the efficient production of electricity while maintaining the waste management benefits of AFBC. A study was undertaken to investigate the technical and economic feasibility of co-firing a PFBC with coal and municipal and industrial wastes. Focus was placed on the production of electricity and the efficient disposal of wastes for application in central power station and distributed locations. Issues concerning waste material preparation and feed, PFBC operation, plant emissions, and regulations are addressed. The results and conclusions developed are generally applicable to current and advanced PFBC design concepts.

  15. STATUS REPORT FOR AGING STUDIES OF EPDM O-RING MATERIAL FOR THE H1616 SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Stefek, T.; Daugherty, W.; Skidmore, E.

    2012-08-31T23:59:59.000Z

    This is an interim status report for tasks carried out per Task Technical Plan SRNL-STI-2011-00506. A series of tasks/experiments are being performed at the Savannah River National Laboratory to monitor the aging performance of ethylene propylene diene monomer (EPDM) Orings used in the H1616 shipping package. The data will support the technical basis to extend the annual maintenance of the EPDM O-rings in the H1616 shipping package and to predict the life of the seals at bounding service conditions. Current expectations are that the O-rings will maintain a seal at bounding normal temperatures in service (152 F) for at least 12 months. The baseline aging data review suggests that the EPDM O-rings are likely to retain significant mechanical properties and sealing force at bounding service temperatures to provide a service life of at least 2 years. At lower, more realistic temperatures, longer service life is likely. Parallel compression stress relaxation and vessel leak test efforts are in progress to further validate this assessment and quantify a more realistic service life prediction. The H1616 shipping package O-rings were evaluated for baseline property data as part of this test program. This was done to provide a basis for comparison of changes in material properties and performance parameters as a function of aging. This initial characterization was limited to physical and mechanical properties, namely hardness, thickness and tensile strength. These properties appear to be consistent with O-ring specifications. Three H1616-1 Containment Vessels were placed in test conditions and are aging at temperatures ranging from 160 to 300 F. The vessels were Helium leak-tested initially and have been tested at periodic intervals after cooling to room temperature to determine if they meet the criterion of leaktightness defined in ANSI standard N14.5-97 (< 1E-07 std cc air/sec at room temperature). To date, no leak test failures have occurred. The cumulative time at temperature ranges from 174 days for the 300 F vessel to 189 days for the 160 F vessel as of 8/1/2012. The compression stress-relaxation (CSR) behavior of H1616 shipping package O-rings is being evaluated to develop an aging model based on material properties. O-ring segments were initially aged at four temperatures (175 F, 235 F, 300 F and 350 F). These temperatures were selected to bound normal service temperatures and to challenge the seals within a reasonable aging period. Currently, samples aging at 300 F and 350 F have reached the mechanical failure point (end of life) which is defined in this study as 90% loss of initial sealing force. As a result, additional samples more recently began aging at {approx}270 F to provide additional data for the aging model. Aging and periodic leak testing of the full containment vessels, as well as CSR testing of O-ring segments is ongoing. Continued testing per the Task Technical Plan is recommended in order to validate the assumptions outlined in this status report and to quantify and validate the long-term performance of O-ring seals under actual service conditions.

  16. Preliminary waste form characteristics report Version 1.0. Revision 1

    SciTech Connect (OSTI)

    Stout, R.B.; Leider, H.R. [eds.

    1991-10-11T23:59:59.000Z

    This report focuses on radioactive waste form characteristics that will be used to design a waste package and an engineered barrier system (EBS) for a suitable repository as part of the Yucca Mountain Project. The term waste form refers to irradiated reactor fuel, other high-level waste (HLW) in various physical forms, and other radioactive materials (other than HLW) which are received for emplacement in a geologic repository. Any encapsulating of stabilizing matrix is also referred to as a waste form.

  17. 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 (OSTI)

    R. A. Carlson

    2005-12-08T23:59:59.000Z

    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.

  18. Remaining Sites Verification Package for the 128-B-2, 100-B Burn Pit #2 Waste Site, Waste Site Reclassification Form 2005-038

    SciTech Connect (OSTI)

    R. A. Carlson

    2005-12-21T23:59:59.000Z

    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.

  19. ENVIROCARE OF UTAH: EXPANDING WASTE ACCEPTANCE CRITERIA TO PROVIDE LOW-LEVEL AND MIXED WASTE DISPOSAL OPTIONS

    SciTech Connect (OSTI)

    Rogers, B.; Loveland, K.

    2003-02-27T23:59:59.000Z

    Envirocare of Utah operates a low-level radioactive waste disposal facility 80 miles west of Salt Lake City in Clive, Utah. Accepted waste types includes NORM, 11e2 byproduct material, Class A low-level waste, and mixed waste. Since 1988, Envirocare has offered disposal options for environmental restoration waste for both government and commercial remediation projects. Annual waste receipts exceed 12 million cubic feet. The waste acceptance criteria (WAC) for the Envirocare facility have significantly expanded to accommodate the changing needs of restoration projects and waste generators since its inception, including acceptable physical waste forms, radiological acceptance criteria, RCRA requirements and treatment capabilities, PCB acceptance, and liquids acceptance. Additionally, there are many packaging, transportation, and waste management options for waste streams acceptable at Envirocare. Many subcontracting vehicles are also available to waste generators for both government and commercial activities.

  20. RH Packaging Operations Manual

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-09-17T23:59:59.000Z

    This procedure provides operating instructions for the RH-TRU 72-B Road Cask, Waste Shipping Package. In this document, ''Packaging'' refers to the assembly of components necessary to ensure compliance with the packaging requirements (not loaded with a payload). ''Package'' refers to a Type B packaging that, with its radioactive contents, is designed to retain the integrity of its containment and shielding when subject to the normal conditions of transport and hypothetical accident test conditions set forth in 10 CFR Part 71. Loading of the RH 72-B cask can be done two ways, on the RH cask trailer in the vertical position or by removing the cask from the trailer and loading it in a facility designed for remote-handling (RH). Before loading the 72-B cask, loading procedures and changes to the loading procedures for the 72-B cask must be sent to CBFO at sitedocuments@wipp.ws for approval.

  1. Final evaluation report for Westinghouse Hanford Company, WRAP-1,208 liter waste drum, docket 94-35-7A, type A packaging

    SciTech Connect (OSTI)

    Kelly, D.L., Westinghouse Hanford

    1996-06-12T23:59:59.000Z

    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.

  2. Production and characterization of a composite insulation material from waste polyethylene teraphtalates

    SciTech Connect (OSTI)

    Kurtulmus, Erhan; Karaboyac?, Mustafa; Yigitarslan, Sibel [Chemical Engineering Department of Suleyman Demirel University, 32200, Isparta (Turkey)

    2013-12-16T23:59:59.000Z

    The pollution of polyethylene teraphtalate (PET) is in huge amounts due to the most widely usage as a packaging material in several industries. Regional pumice has several desirable characteristics such as porous structure, low-cost and light-weight. Considering the requirements approved by the Ministry of Public Works on isolation, composite insulation material consisting of PET and pumice was studied. Sheets of composites differing both in particle size of pumice and composition of polymer were produced by hot-molding technique. Characterization of new composite material was achieved by measuring its weight, density, flammability, endurance against both to common acids and bases, and to a force applied, heat insulation and water adsorption capacity. The results of the study showed that produced composite material is an alternative building material due to its desirable characteristics; low weight, capability of low heat conduction.

  3. Material and energy recovery in integrated waste management systems: A life-cycle costing approach

    SciTech Connect (OSTI)

    Massarutto, Antonio [University of Udine, Udine (Italy); IEFE, Bocconi University, Milan (Italy); Carli, Alessandro de, E-mail: alessandro.decarli@unibocconi.it [IEFE, Bocconi University, Milan (Italy); Graffi, Matteo [University of Udine, Udine (Italy); IEFE, Bocconi University, Milan (Italy)

    2011-09-15T23:59:59.000Z

    Highlights: > The study aims at assessing economic performance of alternative scenarios of MSW. > The approach is the life-cycle costing (LCC). > Waste technologies must be considered as complementary into an integrated strategy. - Abstract: A critical assumption of studies assessing comparatively waste management options concerns the constant average cost for selective collection regardless the source separation level (SSL) reached, and the neglect of the mass constraint. The present study compares alternative waste management scenarios through the development of a desktop model that tries to remove the above assumption. Several alternative scenarios based on different combinations of energy and materials recovery are applied to two imaginary areas modelled in order to represent a typical Northern Italian setting. External costs and benefits implied by scenarios are also considered. Scenarios are compared on the base of the full cost for treating the total waste generated in the area. The model investigates the factors that influence the relative convenience of alternative scenarios.

  4. Level 1 2013/14 Materials Engineering

    E-Print Network [OSTI]

    Harman, Neal.A.

    Materials Science and Engineering (Enhanced with year in industry)[J503] Coordinator: Dr. M Evans Semester 1 will be published online. #12;Module Content: Environment Pollution (Air, water, ground, noise) Bio ­ diversity) Construction Waste Concrete and low carbon alternatives Packaging Waste Climate Change Energy (Electricity

  5. Materials testing for in situ stabilization treatability study of INEEL mixed wastes soils

    SciTech Connect (OSTI)

    Heiser, J.; Fuhrmann, M. [Brookhaven National Lab., Upton, NY (United States)

    1997-09-01T23:59:59.000Z

    This report describes the contaminant-specific materials testing phase of the In Situ Stabilization Comprehensive Environment Response, Compensation, and Liability Act (CERCLA) Treatability Study (TS). The purpose of materials testing is to measure the effectiveness of grouting agents to stabilize Idaho National Engineering and Environmental Laboratory (INEEL) Acid Pit soils and select a grout material for use in the Cold Test Demonstration and Acid Pit Stabilization Treatability Study within the Subsurface Disposal Area (SDA) at the Radioactive Waste Management Complex (RWMC). Test results will assist the selecting a grout material for the follow-on demonstrations described in Test Plan for the Cold Test Demonstration and Acid Pit Stabilization Phases of the In Situ Stabilization Treatability Study at the Radioactive Waste Management Complex.

  6. Evaluation of glass-contact materials for waste glass melters. [Corhart 1215, Corhart 1780, ER-2161

    SciTech Connect (OSTI)

    Rankin, W.N.

    1983-01-01T23:59:59.000Z

    Alternative refractory and electrode materials have been evaluated as glass-contact materials for the vitrification of SRP waste. Monofrax K-3 (The Carborundum Co.) is the optimum refractory, based on corrosion and thermal-shock resistance. Inconel 690 (Huntington Alloys, Inc.) is the recommended electrode alloy. However, a potentially more corrosion-resistant alternative, UCAR Metal Ceramic, Grade LT-1 (Union Carbide Corp.), is being evaluated further. This material melts at a much higher temperature than Inconel 690. In the event of an unexpected high-temperature excursion in a melter, this material may offer significantly greater protection.

  7. Assessment and evaluation of a safety factor with respect to ocean disposal of waste materials

    E-Print Network [OSTI]

    Zapatka, Thomas Francis

    1976-01-01T23:59:59.000Z

    &M University Chairman of Advisory Committee: Dr. Roy W. Harm, Jr. In the past, waste materials too toxic or hazardous for disposal on land or in estuaries have customarily been disposed of at sea. Assessment of the risk and consequent environmental harm... was to investigate and define the correlations between acute and chronic toxicities for known hazardous materials. This relationship is vital for the deter- mination of an appropriate safety factor to be used in evaluating allowable discharges or concentrations...

  8. Assessment and evaluation of a safety factor with respect to ocean disposal of waste materials 

    E-Print Network [OSTI]

    Zapatka, Thomas Francis

    1976-01-01T23:59:59.000Z

    &M University Chairman of Advisory Committee: Dr. Roy W. Harm, Jr. In the past, waste materials too toxic or hazardous for disposal on land or in estuaries have customarily been disposed of at sea. Assessment of the risk and consequent environmental harm... was to investigate and define the correlations between acute and chronic toxicities for known hazardous materials. This relationship is vital for the deter- mination of an appropriate safety factor to be used in evaluating allowable discharges or concentrations...

  9. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    SciTech Connect (OSTI)

    NONE

    1994-10-01T23:59:59.000Z

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G{sub eff}, is defined by: G{sub eff} - {Sigma}{sub M} (F{sub M} x G{sub M}) F{sub M}-fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials.

  10. Seal welded cast iron nuclear waste container

    DOE Patents [OSTI]

    Filippi, Arthur M. (Pittsburgh, PA); Sprecace, Richard P. (Murrysville, PA)

    1987-01-01T23:59:59.000Z

    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.

  11. Remediation of AMD using natural and waste material

    SciTech Connect (OSTI)

    Basir, Nur Athirah Mohamad; Yaacob, Wan Zuhairi Wan [Pusat pengajian Sains Sekitaran dan Sumber Alam, Fakulti Sains dan Teknologi, Universiti Kebangsaan (Malaysia)

    2014-09-03T23:59:59.000Z

    Acid Mine Drainage (AMD) is highly acidic, sulphate rich and frequently carries a high transition metal and heavy metal burden. These AMD's eventually migrate into streams and rivers and impact negatively on the quality of these water bodies. So it is dire necessary to treat this AMD. Various materials such as ladle furnace slag (LFS), bentonite, zeolite, active carbon and kaolinite are currently available to remove heavy metals from contaminated water. All these materials are capable to rise up the pH value and adsorb heavy metals. The process is divided into two stages; screening test and tank experiment. Screening test is conduct by using Batch Equilibrium Test (BET), X-Ray Fluorescene (XRF) identification also Scanning Electron Microscopic (SEM) characteristic. The results showed that all the concentration of heavy metal are decreasing extremely and pH value rise up except for kaolinite. From screening test only ladle furnace slag, bentonite, zeolite and active carbon are chosen for the tank experiment. Tank experiment design with 18cm (H) X 15cm (L) X 15cm (H) was made by silica glass. All these treatment materials were stirred in the tank for 30 days. Initial pH for all tanks is 2.4 and after 30 days is changing into 6.11, 3.91, 2.98 and 2.71 for LFS, bentonite, active carbon as well as zeolite respectively. LFS is the best material for absorption of Zn, Mn and Cu in the synthetic solution. Meanwhile, bentonite is the best absorbent for Ni, Fe and Cd. The conclusion shows that LFS might have big potentials to control AMD pollution base on neutralize pH resulting in a great improvement in the quality of the water.

  12. Packaged die heater

    DOE Patents [OSTI]

    Spielberger, Richard; Ohme, Bruce Walker; Jensen, Ronald J.

    2011-06-21T23:59:59.000Z

    A heater for heating packaged die for burn-in and heat testing is described. The heater may be a ceramic-type heater with a metal filament. The heater may be incorporated into the integrated circuit package as an additional ceramic layer of the package, or may be an external heater placed in contact with the package to heat the die. Many different types of integrated circuit packages may be accommodated. The method provides increased energy efficiency for heating the die while reducing temperature stresses on testing equipment. The method allows the use of multiple heaters to heat die to different temperatures. Faulty die may be heated to weaken die attach material to facilitate removal of the die. The heater filament or a separate temperature thermistor located in the package may be used to accurately measure die temperature.

  13. DOE Order Self Study Modules - DOE O 460.1C Packaging and Transportation Safety and DOE O 460.2A Departmental Materials Transportation and Packaging Management

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,Office of Policy, OAPM |TRU WasteAdministrator |20.1C Briefing

  14. Magnetic mesoporous materials for removal of environmental wastes

    SciTech Connect (OSTI)

    Kim, Byoung Chan; Lee, Jinwoo; Um, Wooyong; Kim, Jaeyun; Joo, Jin; Lee, Jin Hyung; Kwak, Ja Hun; Kim, Jae Hyun; Lee, Changha; Lee, Hongshin; Addleman, Raymond S.; Hyeon, Taeghwan; Gu, Man Bock; Kim, Jungbae

    2011-09-15T23:59:59.000Z

    We have synthesized two different magnetic mesoporous materials that can be easily separated from aqueous solutions by applying a magnetic field. Synthesized magnetic mesoporous materials, Mag-SBA-15 (magnetic ordered mesoporous silica) and Mag-OMC (magnetic ordered mesoporous carbon), have a high loading capacity of contaminants due to high surface area of the supports and high magnetic activity due to the embedded iron oxide particles. Application of surface-modified Mag-SBA-15 was investigated for the collection of mercury from water. The mercury adsorption using Mag-SBA-15 was rapid during the initial contact time and reached a steady-state condition, with an uptake of approximately 97% after 7 hours. Application of Mag-OMC for collection of organics from water, using fluorescein as an easily trackable model analyte, was explored. The fluorescein was absorbed into Mag-OMC within minutes and the fluorescent intensity of solution was completely disappeared after an hour. In another application, Mag-SBA-15 was used as a host of tyrosinase, and employed as recyclable catalytic scaffolds for tyrosinase-catalyzed biodegradation of catechol. Tyrosinase aggregates in Mag-SBA-15, prepared in a two step process of tyrosinase adsorption and crosslinking, could be used repeatedly for catechol degradation with no serious loss of enzyme activity. Considering these results of cleaning up water from toxic inorganic, organic and biochemical contaminants, magnetic mesoporous materials have a great potential to be employed for the removal of environmental contaminants and potentially for the application in large-scale wastewater treatment plants.

  15. Overview on backfill materials and permeable reactive barriers for nuclear waste disposal facilities.

    SciTech Connect (OSTI)

    Moore, Robert Charles; Hasan, Ahmed Ali Mohamed; Holt, Kathleen Caroline; Hasan, Mahmoud A. (Egyptian Atomic Energy Authority, Cairo, Egypt)

    2003-10-01T23:59:59.000Z

    A great deal of money and effort has been spent on environmental restoration during the past several decades. Significant progress has been made on improving air quality, cleaning up and preventing leaching from dumps and landfills, and improving surface water quality. However, significant challenges still exist in all of these areas. Among the more difficult and expensive environmental problems, and often the primary factor limiting closure of contaminated sites following surface restoration, is contamination of ground water. The most common technology used for remediating ground water is surface treatment where the water is pumped to the surface, treated and pumped back into the ground or released at a nearby river or lake. Although still useful for certain remediation scenarios, the limitations of pump-and-treat technologies have recently been recognized, along with the need for innovative solutions to ground-water contamination. Even with the current challenges we face there is a strong need to create geological repository systems for dispose of radioactive wastes containing long-lived radionuclides. The potential contamination of groundwater is a major factor in selection of a radioactive waste disposal site, design of the facility, future scenarios such as human intrusion into the repository and possible need for retrieving the radioactive material, and the use of backfills designed to keep the radionuclides immobile. One of the most promising technologies for remediation of contaminated sites and design of radioactive waste repositories is the use of permeable reactive barriers (PRBs). PRBs are constructed of reactive material(s) to intercept and remove the radionuclides from the water and decontaminate the plumes in situ. The concept of PRBs is relatively simple. The reactive material(s) is placed in the subsurface between the waste or contaminated area and the groundwater. Reactive materials used thus far in practice and research include zero valent iron, hydroxyapatite, magnesium oxide, and others. As the contaminant moves through the reactive material, the contaminant is either sorbed by the reactive material or chemically reacts with the material to form a less harmful substance. Because of the high risk associated with failure of a geological repository for nuclear waste, most nations favor a near-field multibarrier engineered system using backfill materials to prevent release of radionuclides into the surrounding groundwater.

  16. Final evaluation & test report for the standard waste box (docket 01-53-7A) type A packaging

    SciTech Connect (OSTI)

    KELLY, D L

    2001-10-15T23:59:59.000Z

    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.

  17. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-09-27T23:59:59.000Z

    Establishes safety requirements for the proper packaging and transportation of offsite shipments and onsite transfers of hazardous materials andor modal transport. Cancels DOE 1540.2 and DOE 5480.3

  18. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-09-27T23:59:59.000Z

    Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Canceled by DOE 460.1A

  19. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-10-02T23:59:59.000Z

    Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1.

  20. Preparation of certified working reference material sources for the national TRU waste performance demonstration program.

    SciTech Connect (OSTI)

    Mecklenburg, S. L. (Sandra L.); Thronas, D. L. (Denise L.); Wong, A. S. (Amy S.); Marshall, Robert S.,; Becker, G. K.

    2003-01-01T23:59:59.000Z

    Traceable non-destructive assay (NDA) standards containing a variety of radionuclides including uranium, americium, and plutonium oxides mixed with an inert matrix were prepared and certified for use in the U .S. Department of Energy's National TRU Waste Program (NTWP) . The NTWP requires traceable nuclear material standards of the Working Reference Material (WRM) class for qualification of NDA instrumentation that is used to quantify nuclear material in DOE-generated waste before the waste is shipped for final disposition at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico . Qualification and approval of measurement systems is accomplished in part through successful participation in the Non-Destructive Assay (NDA) Performance Demonstration Program (PDP) and is required for DOE and EPA regulatory compliance . An overview of the PDP program highlighting the role of the certified WRMs fabricated at LANL is presented, as well as a summary of the WRM fabrication process and an overview of the inventory of over 175 WRMs fabricated and deployed to DOE measurement facilities to date .

  1. ONGOING INVESTIGATION OF THE EFFECT THAT DRUGSTORE BEETLES HAVE ON CELOTEX ASSEMBLIES FOUND WITHIN RADIOACTIVE MATERIAL PACKAGINGS

    SciTech Connect (OSTI)

    Loftin, B.

    2009-06-08T23:59:59.000Z

    During normal operations at the Department of Energy's Hanford Site in Hanford, WA, drugstore beetles were found within the fiberboard subassemblies of two 9975 Shipping Packages. The Department of Energy's Packaging Certification Program (EM-60) directed a thorough investigation to determine if the drugstore beetles were causing damage that would be detrimental to the safety performance of the Celotex. The Savannah River National Laboratory is continuing to conduct the investigation with entomological expertise being provided by Clemson University. The outcome from the investigation conducted over the previous year was that no discernible damage had been caused by the drugstore beetles. One of the two packages has been essentially untouched over the past year and has only been opened to visually inspect for additional damage. This paper will provide details and results of the ongoing investigation of that package.

  2. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    SciTech Connect (OSTI)

    Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-08-01T23:59:59.000Z

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

  3. Mixing of incompatible materials in waste tanks technical basis document

    SciTech Connect (OSTI)

    SANDGREN, K.R.

    2003-03-21T23:59:59.000Z

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSCs) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report.

  4. Optimal segmentation and packaging process

    DOE Patents [OSTI]

    Kostelnik, Kevin M. (Idaho Falls, ID); Meservey, Richard H. (Idaho Falls, ID); Landon, Mark D. (Idaho Falls, ID)

    1999-01-01T23:59:59.000Z

    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.

  5. Plutonium stabilization and packaging system

    SciTech Connect (OSTI)

    NONE

    1996-05-01T23:59:59.000Z

    This document describes the functional design of the Plutonium Stabilization and Packaging System (Pu SPS). The objective of this system is to stabilize and package plutonium metals and oxides of greater than 50% wt, as well as other selected isotopes, in accordance with the requirements of the DOE standard for safe storage of these materials for 50 years. This system will support completion of stabilization and packaging campaigns of the inventory at a number of affected sites before the year 2002. The package will be standard for all sites and will provide a minimum of two uncontaminated, organics free confinement barriers for the packaged material.

  6. Remaining Sites Verification Package for the 1607-F3 Sanitary Sewer System, Waste Site Reclassification Form 2006-047

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-04-26T23:59:59.000Z

    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. Methods of chemical analysis for organic waste constituents in radioactive materials: A literature review

    SciTech Connect (OSTI)

    Clauss, S.A.; Bean, R.M.

    1993-02-01T23:59:59.000Z

    Most of the waste generated during the production of defense materials at Hanford is presently stored in 177 underground tanks. Because of the many waste treatment processes used at Hanford, the operations conducted to move and consolidate the waste, and the long-term storage conditions at elevated temperatures and radiolytic conditions, little is known about most of the organic constituents in the tanks. Organics are a factor in the production of hydrogen from storage tank 101-SY and represent an unresolved safety question in the case of tanks containing high organic carbon content. In preparation for activities that will lead to the characterization of organic components in Hanford waste storage tanks, a thorough search of the literature has been conducted to identify those procedures that have been found useful for identifying and quantifying organic components in radioactive matrices. The information is to be used in the planning of method development activities needed to characterize the organics in tank wastes and will prevent duplication of effort in the development of needed methods.

  8. Material and energy recovery in integrated waste management systems: The potential for energy recovery

    SciTech Connect (OSTI)

    Consonni, Stefano [Department of Energy, Politecnico di Milano, Via Lambruschini 4, 20156 Milan (Italy); LEAP - Laboratorio Energia Ambiente Piacenza, Via Bixio 27, 29100 Piacenza (Italy); Vigano, Federico, E-mail: federico.vigano@polimi.it [Department of Energy, Politecnico di Milano, Via Lambruschini 4, 20156 Milan (Italy); LEAP -Laboratorio Energia Ambiente Piacenza, Via Bixio 27, 29100 Piacenza (Italy)

    2011-09-15T23:59:59.000Z

    Highlights: > The amount of waste available for energy recovery is significantly higher than the Unsorted Residual Waste (URW). > Its energy potential is always higher than the complement to 100% of the Source Separation Level (SSL). > Increasing SSL has marginal effects on the potential for energy recovery. > Variations in the composition of the waste fed to WtE plants affect only marginally their performances. > A large WtE plant with a treatment capacity some times higher than a small plant achieves electric efficiency appreciably higher. - Abstract: This article is part of a set of six coordinated papers reporting the main findings of a research project carried out by five Italian universities on 'Material and energy recovery in Integrated Waste Management Systems (IWMS)'. An overview of the project and a summary of the most relevant results can be found in the introductory article of the series. This paper describes the work related to the evaluation of mass and energy balances, which has consisted of three major efforts (i) development of a model for quantifying the energy content and the elemental compositions of the waste streams appearing in a IWMS; (ii) upgrade of an earlier model to predict the performances of Waste-to-Energy (WtE) plants; (iii) evaluation of mass and energy balances of all the scenarios and the recovery paths considered in the project. Results show that not only the amount of material available for energy recovery is significantly higher than the Unsorted Residual Waste (URW) left after Separate Collection (SC), because selection and recycling generate significant amounts of residues, but its heating value is higher than that of the original, gross waste. Therefore, the energy potential of what is left after recycling is always higher than the complement to 100% of the Source Separation Level (SSL). Also, increasing SSL has marginal effects on the potential for energy recovery: nearly doubling SSL (from 35% to 65%) reduces the energy potential only by one fourth. Consequently, even at high SSL energy recovery is a fundamental step of a sustainable waste management system. Variations of SSL do bring about variations of the composition, heating value and moisture content of the material fed to WtE plants, but these variations (i) are smaller than one can expect; (ii) have marginal effects on the performances of the WtE plant. These considerations suggest that the mere value of SSL is not a good indicator of the quality of the waste management system, nor of its energy and environmental outcome. Given the well-known dependence of the efficiency of steam power plants with their power output, the efficiency of energy recovery crucially depends on the size of the IWMS served by the WtE plant. A fivefold increase of the amount of gross waste handled in the IWMS (from 150,000 to 750,000 tons per year of gross waste) allows increasing the electric efficiencies of the WtE plant by about 6-7 percentage points (from 21-23% to 28.5% circa).

  9. Appraisal of a cementitious material for waste disposal: Neutron imaging studies of pore structure and sorptivity

    SciTech Connect (OSTI)

    McGlinn, Peter J., E-mail: pjm@ansto.gov.a [Australian Nuclear Science and Technology Organisation, New Illawarra Road, Lucas Heights, NSW 2234 (Australia); Beer, Frikkie C. de [South African Nuclear Energy Corporation (Necsa), Church Street West Extension, Pelindaba, Brits District, Pretoria 0001 (South Africa); Aldridge, Laurence P. [Australian Nuclear Science and Technology Organisation, New Illawarra Road, Lucas Heights, NSW 2234 (Australia); Radebe, Mabuti J.; Nshimirimana, Robert [South African Nuclear Energy Corporation (Necsa), Church Street West Extension, Pelindaba, Brits District, Pretoria 0001 (South Africa); Brew, Daniel R.M.; Payne, Timothy E.; Olufson, Kylie P. [Australian Nuclear Science and Technology Organisation, New Illawarra Road, Lucas Heights, NSW 2234 (Australia)

    2010-08-15T23:59:59.000Z

    Cementitious materials are conventionally used in conditioning intermediate and low level radioactive waste. In this study a candidate cement-based wasteform has been investigated using neutron imaging to characterise the wasteform for disposal in a repository for radioactive materials. Imaging showed both the pore size distribution and the extent of the cracking that had occurred in the samples. The rate of the water penetration measured both by conventional sorptivity measurements and neutron imaging was greater than in pastes made from Ordinary Portland Cement. The ability of the cracks to distribute the water through the sample in a very short time was also evident. The study highlights the significant potential of neutron imaging in the investigation of cementitious materials. The technique has the advantage of visualising and measuring, non-destructively, material distribution within macroscopic samples and is particularly useful in defining movement of water through the cementitious materials.

  10. Networks of recyclable material waste-picker’s cooperatives: An alternative for the solid waste management in the city of Rio de Janeiro

    SciTech Connect (OSTI)

    Tirado-Soto, Magda Martina, E-mail: magda@pep.ufrj.br [Program of Production Engineering, School and Research in Engineering, Federal University of Rio de Janeiro (Brazil); Zamberlan, Fabio Luiz, E-mail: fabio@pep.ufrj.br [Program of Production Engineering, School and Research in Engineering, Federal University of Rio de Janeiro (Brazil)

    2013-04-15T23:59:59.000Z

    Highlights: ? In the marketing of recyclable materials, the waste-pickers are the least wins. ? It is proposed creating a network of recycling cooperatives to achieve viability. ? The waste-pickers contribute to waste management to the city. - Abstract: The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city’s main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers’ cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro.

  11. Radioactive Waste Radioactive Waste

    E-Print Network [OSTI]

    Slatton, Clint

    form · Separate liquid from solid · Radionuclide · Separate all but H3/C14 #12;#12;Radioactive Waste;Radioactive Waste H3/C14 solids Type B (non-incinerable) metal glass hazardous materials #12;#12;Radioactive#12;Radioactive Waste at UF Bldg 831 392-8400 #12;Radioactive Waste · Program is designed to

  12. Method and system including a double rotary kiln pyrolysis or gasification of waste material

    DOE Patents [OSTI]

    McIntosh, M.J.; Arzoumanidis, G.G.

    1997-09-02T23:59:59.000Z

    A method is described for destructively distilling an organic material in particulate form wherein the particulates are introduced through an inlet into one end of an inner rotating kiln ganged to and coaxial with an outer rotating kiln. The inner and outer kilns define a cylindrical annular space with the inlet being positioned in registry with the axis of rotation of the ganged kilns. During operation, the temperature of the wall of the inner rotary kiln at the inlet is not less than about 500 C to heat the particulate material to a temperature in the range of from about 200 C to about 900 C in a pyrolyzing atmosphere to reduce the particulate material as it moves from the one end toward the other end. The reduced particulates including char are transferred to the annular space between the inner and the outer rotating kilns near the other end of the inner rotating kiln and moved longitudinally in the annular space from near the other end toward the one end in the presence of oxygen to combust the char at an elevated temperature to produce a waste material including ash. Also, heat is provided which is transferred to the inner kiln. The waste material including ash leaves the outer rotating kiln near the one end and the pyrolysis vapor leaves through the particulate material inlet. 5 figs.

  13. Method and system including a double rotary kiln pyrolysis or gasification of waste material

    DOE Patents [OSTI]

    McIntosh, Michael J. (Bolingbrook, IL); Arzoumanidis, Gregory G. (Naperville, IL)

    1997-01-01T23:59:59.000Z

    A method of destructively distilling an organic material in particulate form wherein the particulates are introduced through an inlet into one end of an inner rotating kiln ganged to and coaxial with an outer rotating kiln. The inner and outer kilns define a cylindrical annular space with the inlet being positioned in registry with the axis of rotation of the ganged kilns. During operation, the temperature of the wall of the inner rotary kiln at the inlet is not less than about 500.degree. C. to heat the particulate material to a temperature in the range of from about 200.degree. C. to about 900.degree. C. in a pyrolyzing atmosphere to reduce the particulate material as it moves from the one end toward the other end. The reduced particulates including char are transferred to the annular space between the inner and the outer rotating kilns near the other end of the inner rotating kiln and moved longitudinally in the annular space from near the other end toward the one end in the presence of oxygen to combust the char at an elevated temperature to produce a waste material including ash. Also, heat is provided which is transferred to the inner kiln. The waste material including ash leaves the outer rotating kiln near the one end and the pyrolysis vapor leaves through the particulate material inlet.

  14. Testing of organic waste surrogate materials in support of the Hanford organic tank program. Final report

    SciTech Connect (OSTI)

    Turner, D.A. [Westinghouse Hanford Co., Richland, WA (United States); Miron, Y. [Bureau of Mines (United States)

    1994-01-01T23:59:59.000Z

    To address safety issues regarding effective waste management efforts of underground organic waste storage tanks at the Hanford Site, the Bureau of Mines conducted a series of tests, at the request of the Westinghouse Hanford company. In this battery of tests, the thermal and explosive characteristics of surrogate materials, chosen by Hanford, were determined. The surrogate materials were mixtures of inorganic and organic sodium salts, representing fuels and oxidants. The oxidants were sodium nitrate and sodium nitrite. The fuels were sodium salts of oxalate, citrate and ethylenediamine tetraacetic acid (EDTA). Polyethylene powder was also used as a fuel with the oxidant(s). Sodium aluminate was used as a diluent. In addition, a sample of FeCN, supplied by Hanford was also investigated.

  15. Savannah River Site Eastern Transportation Hub: A Concept For a DOE Eastern Packaging, Staging and Maintenance Center - 13143

    SciTech Connect (OSTI)

    England, Jeffery L. [Savannah River National Laboratory, Aiken, South Carolina (United States)] [Savannah River National Laboratory, Aiken, South Carolina (United States); Adams, Karen; Maxted, Maxcine; Ruff Jr, Clarence [U.S. Department of Energy, Savannah River Site, Aiken, SC (United States)] [U.S. Department of Energy, Savannah River Site, Aiken, SC (United States); Albenesius, Andrew; Bowers, Mark D.; Fountain, Geoffrey; Hughes, Michael [Savannah River Nuclear Solutions, Aiken, SC (United States)] [Savannah River Nuclear Solutions, Aiken, SC (United States); Gordon, Sydney [National Security Technologies, LLC, Las Vegas, NV (United States)] [National Security Technologies, LLC, Las Vegas, NV (United States); O'Connor, Stephen [U.S. Department of Energy, HQ DOE, EM-33, Germantown MD (United States)] [U.S. Department of Energy, HQ DOE, EM-33, Germantown MD (United States)

    2013-07-01T23:59:59.000Z

    The Department of Energy (DOE) is working to de-inventory sites and consolidate hazardous materials for processing and disposal. The DOE administers a wide range of certified shipping packages for the transport of hazardous materials to include Special Nuclear Material (SNM), radioactive materials, sealed sources and radioactive wastes. A critical element to successful and safe transportation of these materials is the availability of certified shipping packages. There are over seven thousand certified packagings (i.e., Type B/Type AF) utilized within the DOE for current missions. The synergistic effects of consolidated maintenance, refurbishment, testing, certification, and costing of these services would allow for efficient management of the packagings inventory and to support anticipated future in-commerce shipping needs. The Savannah River Site (SRS) receives and ships radioactive materials (including SNM) and waste on a regular basis for critical missions such as consolidated storage, stabilization, purification, or disposition using H-Canyon and HB-Line. The Savannah River National Laboratory (SRNL) has the technical capability and equipment for all aspects of packaging management. SRS has the only active material processing facility in the DOE complex and is one of the sites of choice for nuclear material consolidation. SRS is a logical location to perform maintenance and periodic testing of the DOE fleet of certified packagings. This initiative envisions a DOE Eastern Packaging Staging and Maintenance Center (PSMC) at the SRS and a western hub at the Nevada National Security Site (NNSS), an active DOE Regional Disposal Site. The PSMC's would be the first place DOE would go to meet their radioactive packaging needs and the primary locations projects would go to disposition excess packaging for beneficial reuse. These two hubs would provide the centralized management of a packaging fleet rather than the current approach to design, procure, maintain and dispose of packagings on a project-by-project basis. This initiative provides significant savings in packaging costs and acceleration of project schedules. In addition to certified packaging, the PSMC would be well suited for select designs of 7A Type A packaging and Industrial Packaging. (authors)

  16. Mixed-layered bismuth-oxygen-iodine materials for capture and waste disposal of radioactive iodine

    DOE Patents [OSTI]

    Krumhansl, James L; Nenoff, Tina M

    2013-02-26T23:59:59.000Z

    Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

  17. Recycled tire rubber and other waste materials in asphalt mixtures. Transportation research record

    SciTech Connect (OSTI)

    NONE

    1995-12-31T23:59:59.000Z

    The papers in this volume, dealing with various facets of recycled tire rubber and other waste materials in asphalt mixtures, should be of interest to state and local construction, design, materials, and research engineers as well as contractors and material producers. In the first papers, Rebala and Estakhri, Malpass and Khosla, and Baker and Connolly describe research related to crumb rubber modified mixtures that was done for the Texas, North Carolina, and New Jersey State Departments of Transportation. Ali et al. report on their research in Canada to determine the feasibility of sing reclaimed roofing materials in hot mix asphalt pavement. Emery discusses the evaluation of 11 Ontario rubber modified demonstration projects in terms of pavement performance, environmental impacts, and recyclability. In the last paper, Fwa and Aziz report on their work in Singapore related to the use of incinerator residue in asphalt mixtures.

  18. Lessons Learned in the Design and Use of IP1 / IP2 Flexible Packaging - 13621

    SciTech Connect (OSTI)

    Sanchez, Mike [VP Global Sales, PacTec, Inc. (United States)] [VP Global Sales, PacTec, Inc. (United States); Reeves, Wendall [National Sales Manager, PacTec, Inc. (United States)] [National Sales Manager, PacTec, Inc. (United States); Smart, Bill [Nuclear Sales Director, PacTec, Inc. (United States)] [Nuclear Sales Director, PacTec, Inc. (United States)

    2013-07-01T23:59:59.000Z

    For many years in the USA, Low Level Radioactive Waste (LLW), contaminated soils and construction debris, have been transported, interim stored, and disposed of, using IP1 / IP2 metal containers. The performance of these containers has been more than adequate, with few safety occurrences. The containers are used under the regulatory oversight of the US Department of Transportation (DOT), 49 Code of Federal Regulations (CFR). In the late 90's the introduction of flexible packaging for the transport, storage, and disposal of low level contaminated soils and construction debris was introduced. The development of flexible packaging came out of a need for a more cost effective package, for the large volumes of waste generated by the decommissioning of many of the US Department of Energy (DOE) legacy sites across the US. Flexible packaging had to be designed to handle a wide array of waste streams, including soil, gravel, construction debris, and fine particulate dust migration. The design also had to meet all of the IP1 requirements under 49CFR 173.410, and be robust enough to pass the IP2 testing 49 CFR 173.465 required for many LLW shipments. Tens of thousands of flexible packages have been safely deployed and used across the US nuclear industry as well as for hazardous non-radioactive applications, with no recorded release of radioactive materials. To ensure that flexible packages are designed properly, the manufacturer must use lessons learned over the years, and the tests performed to provide evidence that these packages are suitable for transporting low level radioactive wastes. The design and testing of flexible packaging for LLW, VLLW and other hazardous waste streams must be as strict and stringent as the design and testing of metal containers. The design should take into consideration the materials being loaded into the package, and should incorporate the right materials, and manufacturing methods, to provide a quality, safe product. Flexible packaging can be shown to meet the criteria for safe and fit for purpose packaging, by meeting the US DOT regulations, and the IAEA Standards for IP-1 and IP-2 including leak tightness. (authors)

  19. Development of New Low-Cost, High-Performance, PV Module Encapsulant/Packaging Materials: Annual Technical Progress Report, Phase 1, 22 October 2002-30 September 2003

    SciTech Connect (OSTI)

    Agro, S. C.; Tucker, R. T.

    2004-03-01T23:59:59.000Z

    The primary objectives of this subcontract are for Specialized Technology Resources, Inc., to work with U.S.-based PV module manufacturers representing crystalline silicon, polycrystalline silicon, amorphous silicon, copper indium diselenide (CIS), and other state-of-the-art thin-film technologies to develop formulations, production processes, prototype and qualify new low-cost, high-performance photovoltaic module encapsulants/packaging materials. The manufacturers will assist in identifying each materials' deficiencies while undergoing development, and then ultimately in qualifying the final optimized materials designed to specifically meet their requirements. Upon completion of this program, new low-cost, high-performance, PV module encapsulant/packaging materials will be qualified, by one or more end-users, for their specific application. Information gathering on topics related to thin-film module technology, including device performance/failure analysis, glass stability, and de vice encapsulation, has been completed. This information has provided concepts and considerations for module failure analysis, accelerated testing design, and encapsulation formulation strategy for thin-film modules.

  20. Geological Problems in Radioactive Waste Isolation: Second Worldwide Review

    E-Print Network [OSTI]

    2010-01-01T23:59:59.000Z

    lived medium level waste (MLW), heat producing vitri- fiedpackage spacing, and waste package heat output, will resultdisposal gallery for heat-emitting waste and to quantify the

  1. Nevada Nuclear Waste Storage Investigations: Exploratory Shaft Facility fluids and materials evaluation

    SciTech Connect (OSTI)

    West, K.A.

    1988-11-01T23:59:59.000Z

    The objective of this study was to determine if any fluids or materials used in the Exploratory Shaft Facility (ESF) of Yucca Mountain will make the mountain unsuitable for future construction of a nuclear waste repository. Yucca Mountain, an area on and adjacent to the Nevada Test Site in southern Nevada, USA, is a candidate site for permanent disposal of high-level radioactive waste from commercial nuclear power and defense nuclear activities. To properly characterize Yucca Mountain, it will be necessary to construct an underground test facility, in which in situ site characterization tests can be conducted. The candidate repository horizon at Yucca Mountain, however, could potentially be compromised by fluids and materials used in the site characterization tests. To minimize this possibility, Los Alamos National Laboratory was directed to evaluate the kinds of fluids and materials that will be used and their potential impacts on the site. A secondary objective was to identify fluids and materials, if any, that should be prohibited from, or controlled in, the underground. 56 refs., 19 figs., 11 tabs.

  2. Landfill Disamenities And Better Utilization of Waste Resources Presented to the Wisconsin Governor's Task Force on Waste Materials Recovery

    E-Print Network [OSTI]

    Columbia University

    're heading, or should be heading regarding solid waste disposal. I began my environmental engineering career in New York State in the 1960's. We had many problems with polluting solid waste dumps, landfill fires, WTE facilities. We know that municipal solid waste, MSW landfills in the US are estimated to release

  3. Recycled Water Reuse Permit Renewal Application for the Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond

    SciTech Connect (OSTI)

    No Name

    2014-10-01T23:59:59.000Z

    ABSTRACT This renewal application for the Industrial Wastewater Reuse Permit (IWRP) WRU-I-0160-01 at Idaho National Laboratory (INL), Materials and Fuels Complex (MFC) Industrial Waste Ditch (IWD) and Industrial Waste Pond (IWP) is being submitted to the State of Idaho, Department of Environmental Quality (DEQ). This application has been prepared in compliance with the requirements in IDAPA 58.01.17, Recycled Water Rules. Information in this application is consistent with the IDAPA 58.01.17 rules, pre-application meeting, and the Guidance for Reclamation and Reuse of Municipal and Industrial Wastewater (September 2007). This application is being submitted using much of the same information contained in the initial permit application, submitted in 2007, and modification, in 2012. There have been no significant changes to the information and operations covered in the existing IWRP. Summary of the monitoring results and operation activity that has occurred since the issuance of the WRP has been included. MFC has operated the IWP and IWD as regulated wastewater land treatment facilities in compliance with the IDAPA 58.01.17 regulations and the IWRP. Industrial wastewater, consisting primarily of continuous discharges of nonhazardous, nonradioactive, routinely discharged noncontact cooling water and steam condensate, periodic discharges of industrial wastewater from the MFC facility process holdup tanks, and precipitation runoff, are discharged to the IWP and IWD system from various MFC facilities. Wastewater goes to the IWP and IWD with a permitted annual flow of up to 17 million gallons/year. All requirements of the IWRP are being met. The Operations and Maintenance Manual for the Industrial Wastewater System will be updated to include any new requirements.

  4. High temperature materials for radioactive waste incineration and vitrification. Revision 1

    SciTech Connect (OSTI)

    Bickford, D F; Ondrejcin, R S; Salley, L

    1986-01-01T23:59:59.000Z

    Incineration or vitrification of radioactive waste subjects equipment to alkaline or acidic fluxing, oxidation, sulfidation, carburization, and thermal shock. It is necessary to select appropriate materials of construction and control operating conditions to avoid rapid equipment failure. Nickel- and cobalt-based alloys with high chromium or aluminum content and aluminum oxide/chromium oxide refractories with high chromium oxide content have provided the best service in pilot-scale melter tests. Inconel 690 and Monofrax K-3 are being used for waste vitrification. Haynes 188 and high alumina refractory are undergoing pilot scale tests for incineration equipment. Laboratory tests indicate that alloys and refractories containing still higher concentrations of chromium or chromium oxide, such as Inconel 671 and Monofrax E, may provide superior resistance to attack in glass melter environments.

  5. Diffusion of Iodine and Rhenium in Category 3 Waste Encasement Concrete and Soil Fill Material

    SciTech Connect (OSTI)

    Wellman, Dawn M.; Mattigod, Shas V.; Whyatt, Greg A.; Powers, Laura; Parker, Kent E.; Wood, Marcus I.

    2006-12-15T23:59:59.000Z

    Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e. sorption or precipitation). This understanding will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. A set of diffusion experiments using carbonated and non-carbonated concrete-soil half cells was conducted under unsaturated conditions (4% and 7% by wt moisture content). Spiked concrete half-cell specimens were prepared with and without colloidal metallic iron addition and were carbonated using supercritical carbon dioxide. Spikes of I and Re were added to achieve measurable diffusion profile in the soil part of the half-cell. In addition, properties of concrete materials likely to influence radionuclide migration such as carbonation were evaluated in an effort to correlate these properties with the release of iodine and rhenium.

  6. Performance-oriented packaging testing of PPP-B-601 ERAPS wood box for packing Group II solid hazardous material. Test report for Oct 91

    SciTech Connect (OSTI)

    Wu, E.

    1991-10-01T23:59:59.000Z

    Qualification tests were performed to determine whether the in-service PPP-B-601 ERAPS Wood Box could be utilized to contain properly dunnaged solid type hazardous materials weighing up to a gross weight of 237 kg (523 pounds). The tests were conducted in accordance with Performance Oriented Packaging (POP) requirements specified by the United Nations Recommendations on the Transportation of Dangerous Goods. The box has conformed to the POP performance requirements; i.e., the box successfully retained its contents throughout the stacking, vibration and drop tests.

  7. Special case waste located at Oak Ridge National Laboratory facilities: Survey report

    SciTech Connect (OSTI)

    Forgy, J.R. Jr.

    1995-11-01T23:59:59.000Z

    Between October 1994 and October 1995, a data base was established at the Oak Ridge National Laboratory (ORNL) to provide a current inventory of the radioactive waste materials, located at ORNL, for which the US Department of Energy (DOE) has no definite planned disposal alternatives. DOE refers to these waste materials as special case waste. To assist ORNL and DOE management in future planning, an inventory system was established and a baseline inventory prepared. This report provides the background of the ORNL special case waste survey project, as well as special case waste category definitions, both current and anticipated sources and locations of special case waste materials, and the survey and data management processes. Special case waste will be that waste material which, no matter how much practical characterization, treatment, and packaging is made, will never meet the acceptance criteria for permanent disposal at ORNL, and does not meet the criteria at a currently planned off-site permanent disposal facility.

  8. WASTE TO WATTS Waste is a Resource!

    E-Print Network [OSTI]

    Columbia University

    to Climate protection in light of the· Waste Framework Directive. The "energy package", e.g. the RenewablesWASTE TO WATTS Waste is a Resource! energy forum Case Studies from Estonia, Switzerland, Germany Bossart,· ABB Waste-to-Energy Plants Edmund Fleck,· ESWET Marcel van Berlo,· Afval Energie Bedrijf From

  9. Removal and recovery of radionuclides and toxic metals from wastes, soils and materials

    SciTech Connect (OSTI)

    Francis, A.J.

    1993-07-01T23:59:59.000Z

    A process has been developed at Brookhaven National Laboratory (BNL) for the removal of metals and radionuclides from contaminated materials, soils, and waste sites (Figure 1). In this process, citric acid, a naturally occurring organic complexing agent, is used to extract metals such as Ba, Cd, Cr, Ni, Zn, and radionuclides Co, Sr, Th, and U from solid wastes by formation of water soluble, metal-citrate complexes. Citric acid forms different types of complexes with the transition metals and actinides, and may involve formation of a bidentate, tridentate, binuclear, or polynuclear complex species. The extract containing radionuclide/metal complex is then subjected to microbiological degradation followed by photochemical degradation under aerobic conditions. Several metal citrate complexes are biodegraded and the metals are recovered in a concentrated form with the bacterial biomass. Uranium forms binuclear complex with citric acid and is not biodegraded. The supernatant containing uranium citrate complex is separated and upon exposure to light, undergoes rapid degradation resulting in the formation of an insoluble, stable polymeric form of uranium. Uranium is recovered as a precipitate (uranium trioxide) in a concentrated form for recycling or for appropriate disposal. This treatment process, unlike others which use caustic reagents, does not create additional hazardous wastes for disposal and causes little damage to soil which can then be returned to normal use.

  10. Assessing recycling versus incineration of key materials in municipal waste: The importance of efficient energy recovery and transport distances

    SciTech Connect (OSTI)

    Merrild, Hanna [Department of Environmental Engineering, Technical University of Denmark, Miljoevej, Building 113, DK-2800 Kongens Lyngby (Denmark); Larsen, Anna W., E-mail: awla@env.dtu.dk [Department of Environmental Engineering, Technical University of Denmark, Miljoevej, Building 113, DK-2800 Kongens Lyngby (Denmark); Christensen, Thomas H. [Department of Environmental Engineering, Technical University of Denmark, Miljoevej, Building 113, DK-2800 Kongens Lyngby (Denmark)

    2012-05-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer We model the environmental impact of recycling and incineration of household waste. Black-Right-Pointing-Pointer Recycling of paper, glass, steel and aluminium is better than incineration. Black-Right-Pointing-Pointer Recycling and incineration of cardboard and plastic can be equally good alternatives. Black-Right-Pointing-Pointer Recyclables can be transported long distances and still have environmental benefits. Black-Right-Pointing-Pointer Paper has a higher environmental benefit than recyclables found in smaller amounts. - Abstract: Recycling of materials from municipal solid waste is commonly considered to be superior to any other waste treatment alternative. For the material fractions with a significant energy content this might not be the case if the treatment alternative is a waste-to-energy plant with high energy recovery rates. The environmental impacts from recycling and from incineration of six material fractions in household waste have been compared through life cycle assessment assuming high-performance technologies for material recycling as well as for waste incineration. The results showed that there are environmental benefits when recycling paper, glass, steel and aluminium instead of incinerating it. For cardboard and plastic the results were more unclear, depending on the level of energy recovery at the incineration plant, the system boundaries chosen and which impact category was in focus. Further, the environmental impact potentials from collection, pre-treatment and transport was compared to the environmental benefit from recycling and this showed that with the right means of transport, recyclables can in most cases be transported long distances. However, the results also showed that recycling of some of the material fractions can only contribute marginally in improving the overall waste management system taking into consideration their limited content in average Danish household waste.

  11. Preliminary Review of the Degradation of Cellulosic, Plastic, and Rubber Materials in the Waste Isolation Pilot Plant, and Possible

    E-Print Network [OSTI]

    Preliminary Review of the Degradation of Cellulosic, Plastic, and Rubber Materials in the Waste............................................... 2-1 2.1.1 Microbial Degradation of Cellulosic, Plastic, and Rubber Materials ...... 2-1 2.1.2 Anoxic-Biodegradable.............................................................................. 3-4 3.2.4 Uncertainties in Cellulosics, Plastics, and Rubber Inventory.................. 3-6 4

  12. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2005-02-28T23:59:59.000Z

    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.

  13. Remaining Sites Verification Package for the 100-B-18, 184-B Powerhouse Debris Pile, Waste Site Reclassification Form 2007-020

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-11-30T23:59:59.000Z

    The 100-B-18 Powerhouse Debris Pile contained miscellaneous demolition waste from the decommissioning activities of the 184-B Powerhouse. The debris covered an area roughly 15 m by 30 m and included materials such as concrete blocks, mixed aggregate/concrete slabs, stone rubble, asphalt rubble, traces of tar/coal, broken fluorescent lights, brick chimney remnants, and rubber hoses. 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.

  14. Packaging laws motivate industry to draw up global recycling pacts

    SciTech Connect (OSTI)

    Kiesche, E.S.

    1992-11-25T23:59:59.000Z

    Spurred by criticism that plastics contribute to the solid waste problem, some plastics industry associations are joining forces to find global solutions. For one, manufacturers of expanded polystyrene (EPS) protective foam packaging from the US, Germany, Austria, and Japan have entered an agreement to encourage the collection and recycling of postconsumer material. Meanwhile, polyvinyl chloride (PVC) makers from Japan, Europe, and North and South America met in September and agreed to develop systems and technology for PVC recycling as well as incineration. The EPS agreement was motivated by fears that packaging material made outside of Germany would not meet the country's new Packaging Ordinance, says a spokesman for the Association of Foam Packaging Recyclers (AFPR; Washington). AFPR chairman John O'Leary Jr., who is also president and CEO of Tuscarora Inc. (New Brighton, PA), says [open quotes]Some consumer product manufacturers in the US incorrectly believe that EPS protective foam packaging may not be acceptable in Germany.[close quotes] The agreement was signed by EPSY-Germany, EPSY-Austria, AFPR, and the Japan EPS Recycling Association. The AFPR is affiliated with the Society of the Plastics Industry (Washington).

  15. Environmental evaluation of municipal waste prevention

    SciTech Connect (OSTI)

    Gentil, Emmanuel C.; Gallo, Daniele [Department of Environmental Engineering, Building 115, Technical University of Denmark, DK-2800 Kongens Lyngby (Denmark); Christensen, Thomas H., E-mail: thho@env.dtu.dk [Department of Environmental Engineering, Building 115, Technical University of Denmark, DK-2800 Kongens Lyngby (Denmark)

    2011-12-15T23:59:59.000Z

    Highlights: > Influence of prevention on waste management systems, excluding avoided production, is relatively minor. > Influence of prevention on overall supply chain, including avoided production is very significant. > Higher relative benefits of prevention are observed in waste management systems relying mainly on landfills. - Abstract: Waste prevention has been addressed in the literature in terms of the social and behavioural aspects, but very little quantitative assessment exists of the environmental benefits. Our study evaluates the environmental consequences of waste prevention on waste management systems and on the wider society, using life-cycle thinking. The partial prevention of unsolicited mail, beverage packaging and food waste is tested for a 'High-tech' waste management system relying on high energy and material recovery and for a 'Low-tech' waste management system with less recycling and relying on landfilling. Prevention of 13% of the waste mass entering the waste management system generates a reduction of loads and savings in the waste management system for the different impacts categories; 45% net reduction for nutrient enrichment and 12% reduction for global warming potential. When expanding our system and including avoided production incurred by the prevention measures, large savings are observed (15-fold improvement for nutrient enrichment and 2-fold for global warming potential). Prevention of food waste has the highest environmental impact saving. Prevention generates relatively higher overall relative benefit for 'Low-tech' systems depending on landfilling. The paper provides clear evidence of the environmental benefits of waste prevention and has specific relevance in climate change mitigation.

  16. Method for making a low density polyethylene waste form for safe disposal of low level radioactive material

    DOE Patents [OSTI]

    Colombo, P.; Kalb, P.D.

    1984-06-05T23:59:59.000Z

    In the method of the invention low density polyethylene pellets are mixed in a predetermined ratio with radioactive particulate material, then the mixture is fed through a screw-type extruder that melts the low density polyethylene under a predetermined pressure and temperature to form a homogeneous matrix that is extruded and separated into solid monolithic waste forms. The solid waste forms are adapted to be safely handled, stored for a short time, and safely disposed of in approved depositories.

  17. NNWSI waste form performance test development

    SciTech Connect (OSTI)

    Bates, J.K.; Gerding, T.J.

    1984-12-31T23:59:59.000Z

    A test method has been developed to measure the release of radionuclides from the waste package under simulated NNWSI repository conditions, and to provide information concerning materials interactions that may occur in the repository. Data from 13 weeks of unsaturated testing are discussed and compared to that from a 13-week analog test. The data indicate that the waste form test is capable of producing consistent, reproducible results that will be useful in evaluating the role of the waste in the long-term performance of the repository. 6 references, 3 figures.

  18. Energy implications of the thermal recovery of biodegradable municipal waste materials in the United Kingdom

    SciTech Connect (OSTI)

    Burnley, Stephen, E-mail: s.j.burnley@open.ac.uk [Open University, Walton Hall, Milton Keynes MK7 6AA (United Kingdom); Phillips, Rhiannon, E-mail: rhiannon.jones@environment-agency.gov.uk [Strategy Unit, Welsh Assembly Government, Ty Cambria, 29 Newport Road, Cardiff CF24 0TP (United Kingdom); Coleman, Terry, E-mail: terry.coleman@erm.com [Environmental Resources Management Ltd, Eaton House, Wallbrook Court, North Hinksey Lane, Oxford OX2 0QS (United Kingdom); Rampling, Terence, E-mail: twa.rampling@hotmail.com [7 Thurlow Close, Old Town Stevenage, Herts SG1 4SD (United Kingdom)

    2011-09-15T23:59:59.000Z

    Highlights: > Energy balances were calculated for the thermal treatment of biodegradable wastes. > For wood and RDF, combustion in dedicated facilities was the best option. > For paper, garden and food wastes and mixed waste incineration was the best option. > For low moisture paper, gasification provided the optimum solution. - Abstract: Waste management policies and legislation in many developed countries call for a reduction in the quantity of biodegradable waste landfilled. Anaerobic digestion, combustion and gasification are options for managing biodegradable waste while generating renewable energy. However, very little research has been carried to establish the overall energy balance of the collection, preparation and energy recovery processes for different types of wastes. Without this information, it is impossible to determine the optimum method for managing a particular waste to recover renewable energy. In this study, energy balances were carried out for the thermal processing of food waste, garden waste, wood, waste paper and the non-recyclable fraction of municipal waste. For all of these wastes, combustion in dedicated facilities or incineration with the municipal waste stream was the most energy-advantageous option. However, we identified a lack of reliable information on the energy consumed in collecting individual wastes and preparing the wastes for thermal processing. There was also little reliable information on the performance and efficiency of anaerobic digestion and gasification facilities for waste.

  19. 2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond

    SciTech Connect (OSTI)

    David B. Frederick

    2011-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000160 01), for the wastewater reuse site at the Idaho National Laboratory Site’s Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond from May 1, 2010 through October 31, 2010. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of special compliance conditions • Discussion of the facility’s environmental impacts During the 2010 partial reporting year, an estimated 3.646 million gallons of wastewater were discharged to the Industrial Waste Ditch and Pond which is well below the permit limit of 13 million gallons per year. The concentrations of all permit-required analytes in the samples from the down gradient monitoring wells were below the Ground Water Quality Rule Primary and Secondary Constituent Standards.

  20. 2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond

    SciTech Connect (OSTI)

    David Frederick

    2012-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA-000160-01), for the wastewater reuse site at the Idaho National Laboratory Site's Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: (1) Facility and system description; (2) Permit required effluent monitoring data and loading rates; (3) Groundwater monitoring data; (4) Status of special compliance conditions; and (5) Discussion of the facility's environmental impacts. During the 2011 reporting year, an estimated 6.99 million gallons of wastewater were discharged to the Industrial Waste Ditch and Pond which is well below the permit limit of 13 million gallons per year. Using the dissolved iron data, the concentrations of all permit-required analytes in the samples from the down gradient monitoring wells were below the Ground Water Quality Rule Primary and Secondary Constituent Standards.

  1. 2013 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2014-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (WRU-I-0160-01, formerly LA 000160 01), for the wastewater reuse site at the Idaho National Laboratory Site’s Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond from November 1, 2012 through October 31, 2013. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of special compliance conditions • Discussion of the facility’s environmental impacts During the 2013 reporting year, an estimated 9.64 million gallons of wastewater were discharged to the Industrial Waste Ditch and Pond which is well below the permit limit of 17 million gallons per year. The concentrations of all permit-required analytes in the samples from the down gradient monitoring wells were below the applicable Idaho Department of Environmental Quality’s groundwater quality standard levels.

  2. 2012 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2013-02-01T23:59:59.000Z

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (WRU-I-0160-01, formerly LA 000160 01), for the wastewater reuse site at the Idaho National Laboratory Site’s Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond from November 1, 2011 through October 31, 2012. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of special compliance conditions • Discussion of the facility’s environmental impacts During the 2012 reporting year, an estimated 11.84 million gallons of wastewater were discharged to the Industrial Waste Ditch and Pond which is well below the permit limit of 17 million gallons per year. The concentrations of all permit-required analytes in the samples from the down gradient monitoring wells were below the Ground Water Quality Rule Primary and Secondary Constituent Standards.

  3. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2003-04-04T23:59:59.000Z

    To establish safety requirements for the proper packaging and transportation of Department of Energy (DOE)/National Nuclear Security Administration (NNSA) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1A. Canceled by DOE O 460.1C.

  4. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-05-14T23:59:59.000Z

    The order establishes safety requirements for the proper packaging and transportation of DOE, including NNSA, offsite shipments and onsite transfers of radioactive and other hazardous materials and for modal transportation. Cancels DOE O 460.1B, 5-14-10

  5. Evidence for the incorporation of lead into barite from waste rock pile materials

    SciTech Connect (OSTI)

    COURTIN-NOMADE, ALEXANDRA; SOUBRAND-COLIN, MARILYNE; MARCUS, MATTHEW A.; FAKRA, SIRINE .C

    2008-01-21T23:59:59.000Z

    Because Pb is one of the most toxic elements and is found as a major contaminant in mining environments, this study aims to identify the distribution of this element in host phases issued from the alteration of mine wastes. The sampling location was a former mine near Oakland, California (USA). This mine was once a source of sulfide minerals from which sulfuric acid was made. The material discussed in this paper was collected in iron hardpans that were formed within the waste rock pile resulting from the excavation work. In most contaminated environments (soils, mine waste), secondary metal-bearing phases arising from alteration processes are usually fine-grained (from 10 {micro}m to less than 1 {micro}m) and highly heterogeneous, requiring the use of micron-scale techniques. We performed micro-Raman spectroscopy, microscanning X-ray diffraction (SXRD), and microextended X-ray near edge spectroscopy (XANES) to determine the relationships between Pb and a Ba/Fe-rich host phase. Micro-Raman spectroscopy suggests that Pb is preferentially incorporated into barite rather than goethite. Results from micro-Raman experiments show the high sensitivity of this analytical tool to the incorporation of Pb into barite by being especially sensitive to the variations of the S-O bond and showing the characteristic bands due to the contribution of Pb. This association is confirmed and is well-illustrated by micro-SXRD mineral species maps showing the correlation between Pb and barite. Microfocused XANES indicates that Pb is present as Pb{sup 2+}, agreeing with the in situ physicochemical parameters.

  6. Waste Management in Dsseldorf Combination of separate collection,

    E-Print Network [OSTI]

    Columbia University

    Waste Management in Düsseldorf Combination of separate collection, recycling and waste-to-energy Biowaste Garden waste Light packaging Paper Glass Wood from bulky waste Bulky waste Rest / mixed waste Bio- Garden- Paper Glass Light Metals Wood Bulky Rest waste waste Card- Pack. waste board Saved CO2

  7. Assessment and recommendations for fissile-material packaging exemptions and general licenses within 10 CFR Part 71

    SciTech Connect (OSTI)

    Parks, C.V.; Hopper, C.M.; Lichtenwalter, J.L. [Oak Ridge National Lab., TN (United States)

    1998-07-01T23:59:59.000Z

    This report provides a technical and regulatory assessment of the fissile material general licenses and fissile material exemptions within Title 10 of the Code of Federal Regulations Part 71. The assessment included literature studies and calculational analyses to evaluate the technical criteria; review of current industry practice and concerns; and a detailed evaluation of the regulatory text for clarity, consistency and relevance. Recommendations for potential consideration by the Nuclear Regulatory Commission staff are provided. The recommendations call for a simplification and consolidation of the general licenses and a change in the technical criteria for the first fissile material exemptions.

  8. 4.0 RISK FROM URANIUM MINING WASTE IN BUILDING In general, building materials contain low levels of radioactivity. For example, the range of

    E-Print Network [OSTI]

    the wastes from uranium mines have been removed from mining sites and used in local and nearby communities4.0 RISK FROM URANIUM MINING WASTE IN BUILDING MATERIALS In general, building materials contain low levels of radioactivity. For example, the range of natural uranium concentrations may average as low

  9. Challenges in the Packaging of MEMS

    SciTech Connect (OSTI)

    Malshe, A.P.; Singh, S.B.; Eaton, W.P.; O'Neal, C.; Brown, W.D.; Miller, W.M.

    1999-03-26T23:59:59.000Z

    The packaging of Micro-Electro-Mechanical Systems (MEMS) is a field of great importance to anyone using or manufacturing sensors, consumer products, or military applications. Currently much work has been done in the design and fabrication of MEMS devices but insufficient research and few publications have been completed on the packaging of these devices. This is despite the fact that packaging is a very large percentage of the total cost of MEMS devices. The main difference between IC packaging and MEMS packaging is that MEMS packaging is almost always application specific and greatly affected by its environment and packaging techniques such as die handling, die attach processes, and lid sealing. Many of these aspects are directly related to the materials used in the packaging processes. MEMS devices that are functional in wafer form can be rendered inoperable after packaging. MEMS dies must be handled only from the chip sides so features on the top surface are not damaged. This eliminates most current die pick-and-place fixtures. Die attach materials are key to MEMS packaging. Using hard die attach solders can create high stresses in the MEMS devices, which can affect their operation greatly. Low-stress epoxies can be high-outgassing, which can also affect device performance. Also, a low modulus die attach can allow the die to move during ultrasonic wirebonding resulting to low wirebond strength. Another source of residual stress is the lid sealing process. Most MEMS based sensors and devices require a hermetically sealed package. This can be done by parallel seam welding the package lid, but at the cost of further induced stress on the die. Another issue of MEMS packaging is the media compatibility of the packaged device. MEMS unlike ICS often interface with their environment, which could be high pressure or corrosive. The main conclusion we can draw about MEMS packaging is that the package affects the performance and reliability of the MEMS devices. There is a gross lack of understanding between the package materials, induced stress, and the device performance. The material properties of these packaging materials are not well defined or understood. Modeling of these materials and processes is far from maturity. Current post-package yields are too low for commercial feasibility, and consumer operating environment reliability and compatibility are often difficult to simulate. With further understanding of the materials properties and behavior of the packaging materials, MEMS applications can be fully realized and integrated into countless commercial and military applications.

  10. Subsolidus sintering of SYNROC: II. Materials selections, process improvements, waste form evaluations

    SciTech Connect (OSTI)

    Palmour, H. III.; Hare, T.M.; Russ, J.C.; Boss, C.B.; Solomah, A.G.; Batchelor, A.D.

    1981-07-01T23:59:59.000Z

    The principal areas of research were related to materials selections and characterization, process optimizations, crystalline phase development, sinterability, resultant microstructures and evaluations of leaching behavior. With and without simulated radwaste doping, the Modified SYNROC-B formulation was found to be sinterable to technical density (D > 0.95 in the CTS mode) at temperatures in the range 1195/sup 0/C to 1285/sup 0/C, depending upon TiO/sub 2/ and CaCO/sub 3/ materials selections, and upon powder processing methods employed prior to firing. Of the 16 TiO/sub 2/ raw materials evaluated in air-fired, undoped batches, 15 yielded technically dense compacts (D > 0.95). Three fine pigmentary grades of TiO/sub 2/ were selected for further study in doped and undoped versions fired in Ar, 4% H/sub 2/. When intensively milled with other well chosen matrix constituents and 10% spray-calcined simulated waste, each of them yielded sintered densities of greater than or equal to 4.2 g/cm/sup 3/ (D greater than or equal to 0.96) at 1260/sup 0/C, 2h in Ar, 4% H/sub 2/ atmosphere. Leachability studies have been carried out in triple distilled H/sub 2/O according to MCC-1 and MCC-2 procedures at 25/sup 0/ and 150/sup 0/C, respectively, and under ..gamma..-irradiation for dose rates of 2-5 x 10/sup 5/ rad/h at approx. 25/sup 0/C. The results obtained showed that freshly exposed interions of sintered Modified SYNROC-B ceramics were highly stable in the leaching environment, and were very retentive of simulated waste ions, including the most leachable species, Cs. Depending on leaching conditions, the highest Cs leach rates (after 3 days) were on the order of 10/sup -1/ g.m/sup -2/.day/sup -1/, but diminished sharply for longer times (up to 92 days) to the range 10/sup -2/ - 10/sup -4/ g.m/sup -2/.day/sup -1/.

  11. Method for contamination control and barrier apparatus with filter for containing waste materials that include dangerous particulate matter

    DOE Patents [OSTI]

    Pinson, P.A.

    1998-02-24T23:59:59.000Z

    A container for hazardous waste materials that includes air or other gas carrying dangerous particulate matter has incorporated barrier material, preferably in the form of a flexible sheet, and one or more filters for the dangerous particulate matter sealably attached to such barrier material. The filter is preferably a HEPA type filter and is preferably chemically bonded to the barrier materials. The filter or filters are preferably flexibly bonded to the barrier material marginally and peripherally of the filter or marginally and peripherally of air or other gas outlet openings in the barrier material, which may be a plastic bag. The filter may be provided with a backing panel of barrier material having an opening or openings for the passage of air or other gas into the filter or filters. Such backing panel is bonded marginally and peripherally thereof to the barrier material or to both it and the filter or filters. A coupling or couplings for deflating and inflating the container may be incorporated. Confining a hazardous waste material in such a container, rapidly deflating the container and disposing of the container, constitutes one aspect of the method of the invention. The chemical bonding procedure for producing the container constitutes another aspect of the method of the invention. 3 figs.

  12. Method for contamination control and barrier apparatus with filter for containing waste materials that include dangerous particulate matter

    DOE Patents [OSTI]

    Pinson, Paul A. (Idaho Falls, ID)

    1998-01-01T23:59:59.000Z

    A container for hazardous waste materials that includes air or other gas carrying dangerous particulate matter has incorporated in barrier material, preferably in the form of a flexible sheet, one or more filters for the dangerous particulate matter sealably attached to such barrier material. The filter is preferably a HEPA type filter and is preferably chemically bonded to the barrier materials. The filter or filters are preferably flexibly bonded to the barrier material marginally and peripherally of the filter or marginally and peripherally of air or other gas outlet openings in the barrier material, which may be a plastic bag. The filter may be provided with a backing panel of barrier material having an opening or openings for the passage of air or other gas into the filter or filters. Such backing panel is bonded marginally and peripherally thereof to the barrier material or to both it and the filter or filters. A coupling or couplings for deflating and inflating the container may be incorporated. Confining a hazardous waste material in such a container, rapidly deflating the container and disposing of the container, constitutes one aspect of the method of the invention. The chemical bonding procedure for producing the container constitutes another aspect of the method of the invention.

  13. Safety evaluation for packaging (onsite) for cesium chloride capsules with type W overpacks

    SciTech Connect (OSTI)

    McCoy, J.C.

    1997-09-15T23:59:59.000Z

    This Safety Evaluation for Packaging (SEP) documents the evaluation of a new basket design and overpacked cesium chloride capsule payload for the Beneficial Uses Shipping System (BUSS) Cask in accordance with the onsite transportation requirements of the Hazardous Material Packaging and Shipping manual, WHC-CM-2-14. This design supports the one-time onsite shipment of 16 cesium chloride capsules with Type W overpacks from the 324 Building to the 224T Building at the Waste Encapsulation and Storage Facility (WESF). The SEP is valid for a one-time onsite shipment or until August 1, 1998, whichever occurs first.

  14. Optimal segmentation and packaging process

    DOE Patents [OSTI]

    Kostelnik, K.M.; Meservey, R.H.; Landon, M.D.

    1999-08-10T23:59:59.000Z

    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.

  15. THE NGA-DOE GRANT TO EXAMINE CRITICAL ISSUES RELATED TO RADIOACTIVE WASTE AND MATERIALS DISPOSITION INVOLVING DOE FACILITIES

    SciTech Connect (OSTI)

    Ann M. Beauchesne

    2000-01-01T23:59:59.000Z

    Through the National Governors Association (NGA) project ``Critical Issues Related to Radioactive Waste and Materials Disposition Involving DOE Facilities'' NGA brings together Governors' policy advisors, state regulators, and DOE officials to examine critical issues related to the cleanup and operation of DOE nuclear weapons and research facilities. Topics explored through this project include: Decisions involving disposal of mixed, low-level, and transuranic (TRU) waste and disposition of nuclear materials; Decisions involving DOE budget requests and their effect on environmental cleanup and compliance at DOE facilities; Strategies to treat mixed, low-level, and transuranic (TRU) waste and their effect on individual sites in the complex; Changes to the FFCA site treatment plans as a result of proposals in the Department's Accelerating Cleanup: Paths to Closure plan and contractor integration analysis; Interstate waste and materials shipments; and Reforms to existing RCRA and CERCLA regulations/guidance to address regulatory overlap and risks posed by DOE wastes. The overarching theme of this project is to help the Department improve coordination of its major program decisions with Governors' offices and state regulators and to ensure such decisions reflect input from these key state officials and stakeholders. This report summarizes activities conducted during the period from October 1, 1999 through January 31, 2000, under the NGA grant. The work accomplished by the NGA project team during the past three months can be categorized as follows: maintained open communication with DOE on a variety of activities and issues within the DOE environmental management complex; convened and facilitated the October 6--8 NGA FFCA Task Force Meeting in Oak Ridge, Tennessee; maintained communication with NGA Federal Facilities Compliance Task Force members regarding DOE efforts to formulate a configuration for mixed low-level waste and low-level treatment and disposal, external regulation of DOE; and continued to facilitate interactions between the states and DOE to develop a foundation for an ongoing substantive relationship between the Governors of key states and the Department.

  16. THE NGA-DOE GRANT TO EXAMINE CRITICAL ISSUES RELATED TO RADIOACTIVE WASTE AND MATERIALS DISPOSITION INVOLVING DOE FACILITIES

    SciTech Connect (OSTI)

    NONE

    1998-07-01T23:59:59.000Z

    Through the National Governors' Association (NGA) project ''Critical Issues Related to Radioactive Waste and Materials Disposition Involving DOE Facilities'' NGA brings together Governors' policy advisors, state regulators, and DOE officials to examine critical issues related to the cleanup and operation of DOE nuclear weapons and research facilities. Topics explored through this project include: Decisions involving disposal of mixed, low-level, and transuranic (TRU) waste and disposition of nuclear materials. Decisions involving DOE budget requests and their effect on environmental cleanup and compliance at DOE facilities. Strategies to treat mixed, low-level, and transuranic (TRU) waste and their effect on individual sites in the complex. Changes to the FFCA site treatment plans as a result of proposals in DOE's Accelerating Cleanup: Paths to Closure strategy and contractor integration analysis. Interstate waste and materials shipments. Reforms to existing RCRA and CERCLA regulations/guidance to address regulatory overlap and risks posed by DOE wastes. The overarching theme of this project is to help the Department improve coordination of its major program decisions with Governors' offices and state regulators and to ensure such decisions reflect input from these key state officials and stakeholders. This report summarizes activities conducted during the quarter from April 30, 1998 through June 30, 1998 under the NGA project. The work accomplished by the NGA project team during the past four months can be categorized as follows: maintained open communication with DOE on a variety of activities and issues within the DOE environmental management complex; and provided ongoing support to state-DOE interactions. maintained communication with NGA Federal Facilities Compliance Task Force members regarding DOE efforts to formulate a configuration for mixed low-level waste and low-level treatment and disposal, DOE's Environmental Management Budget, and DOE's proposed Intersite Discussions.

  17. Transuranic contaminated waste container characterization and data base. Revision I

    SciTech Connect (OSTI)

    Kniazewycz, B.G.

    1980-05-01T23:59:59.000Z

    The Nuclear Regulatory Commission (NRC) is developing regulations governing the management, handling and disposal of transuranium (TRU) radioisotope contaminated wastes as part of the NRC's overall waste management program. In the development of such regulations, numerous subtasks have been identified which require completion before meaningful regulations can be proposed, their impact evaluated and the regulations implemented. This report was prepared to assist in the development of the technical data base necessary to support rule-making actions dealing with TRU-contaminated wastes. An earlier report presented the waste sources, characteristics and inventory of both Department of Energy (DOE) generated and commercially generated TRU waste. In this report a wide variety of waste sources as well as a large TRU inventory were identified. The purpose of this report is to identify the different packaging systems used and proposed for TRU waste and to document their characteristics. This document then serves as part of the data base necessary to complete preparation and initiate implementation of TRU waste container and packaging standards and criteria suitable for inclusion in the present TRU waste management program. It is the purpose of this report to serve as a working document which will be used as appropriate in the TRU Waste Management Program. This report, and those following, will be compatible not only in format, but also in reference material and direction.

  18. Apparatus and method for quantitative assay of samples of transuranic waste contained in barrels in the presence of matrix material

    DOE Patents [OSTI]

    Caldwell, J.T.; Herrera, G.C.; Hastings, R.D.; Shunk, E.R.; Kunz, W.E.

    1987-08-28T23:59:59.000Z

    Apparatus and method for performing corrections for matrix material effects on the neutron measurements generated from analysis of transuranic waste drums using the differential-dieaway technique. By measuring the absorption index and the moderator index for a particular drum, correction factors can be determined for the effects of matrix materials on the ''observed'' quantity of fissile and fertile material present therein in order to determine the actual assays thereof. A barrel flux monitor is introduced into the measurement chamber to accomplish these measurements as a new contribution to the differential-dieaway technology. 9 figs.

  19. Packaging and Transportation Support at LANL CTMA 2012

    SciTech Connect (OSTI)

    Salazar, Nick [Los Alamos National Laboratory

    2012-06-08T23:59:59.000Z

    Operations Support Packaging and Transportation (OS-PT) supports LANL in various functions. Some highlights of the past year have been with the work relating to environmental remediation, type B packaging, non-DOT compliant transfers, and special permit training. The TA-21 remediation project was part of the ARRA funding that LANL received. The $212 million in funding was used to demolish 24 buildings at TA-21, excavate the lab's oldest waste disposal site, and install 16 groundwater monitoring wells. The project was completed ahead of schedule and under budget. More than 300 tons of metal was recycled and all the soil excavated from MDA-B was replaced with clean fill. OS-PT supported this projected by transporting more than 7 million pounds of waste to TA-54 Area G with an addendum to their TSD. Because of the public access on the transfer route, Los Alamos County restricted the transfer to happen from 2:00 AM to 4:00 AM. OS-PT conducted 8 transfers in support of this project. Some concerns included the contaminated trailers at receipt facilities when transferring filled Super Sacks. Future Super Sacks were over packed into new IP-2 Super Sacks before shipping. OS-PT is also supporting the remediation of TA-54 Area G. LANL has an agreement with the State of New Mexico to remove all TRU waste currently stored above ground from at Area G. OS-PT supports this initiative with transfers of TRU waste under LANL's TSD and support of TRU shipments to WIPP. Another project supported by our organization is gas cylinder/dewar recycling and remediation. We are focusing on reducing risk associated with unneeded gasses at LANL. To minimized excessive ordering, to save money and time, and to minimize hazards OS-PT is supporting a gas recycling program. This program will allow programmatic organization across LANL to share unused/unneeded gasses. Instead of old dewars being disposed of, OS-PT has began identifying these dewars and sending them for refurbishment. To date, this effort has saved LANL $450K and estimated saving for future efforts will be more than $1.5 million. Some Projects that are happening here at LANL are offsite source recovery, weapon component transfers, and isotope science production. There are specific packages that help support these projects for the shipment of related materials. OS-PT provides support to these packages to ensure they are and will be available to continue this support. The Areva 435-B Overpack will help the Offsite Source Recovery Project recover high activity gamma sources from various locations across the globe. The Safety Analysis for Packaging is scheduled for initial completion June of 2012. The DPP-1 package is designed to replace the Model FL, which was designed by Rocky Flats and began service in 1990. LANL has collaborated on package design with LLNL, Pantex, Y-12, and KCP. LANL is supporting LLNL on component fixture development. Testing to 10 CFR 71 is to be completed in the Fall of 2012 and scheduled for NA-174 approval in 2014. The SAFESHIELD package helps supports LANL's Isotope production projects. This package can transfer highly irradiated materials from LANL's accelerator to material processing facilities. LANL worked to renew the SAFESHEILD's Certification for 5 more years.

  20. Ceremony Theming Packages Package One

    E-Print Network [OSTI]

    Tobar, Michael

    will incur a surcharge of $50.00 per additional half hour Weddings at The University of Western Australia #12: _____________________________________________________________________________________________________ Weddings at The University of Western Australia #12;Ceremony Theming Packages Payment Details CREDIT CARD I, ______________________________________, authorise The University Club of Western Australia P/L, to debit the amount

  1. Waste management facilities cost information for transportation of radioactive and hazardous materials

    SciTech Connect (OSTI)

    Feizollahi, F.; Shropshire, D.; Burton, D.

    1995-06-01T23:59:59.000Z

    This report contains cost information on the U.S. Department of Energy (DOE) Complex waste streams that will be addressed by DOE in the programmatic environmental impact statement (PEIS) project. It describes the results of the task commissioned by DOE to develop cost information for transportation of radioactive and hazardous waste. It contains transportation costs for most types of DOE waste streams: low-level waste (LLW), mixed low-level waste (MLLW), alpha LLW and alpha MLLW, Greater-Than-Class C (GTCC) LLW and DOE equivalent waste, transuranic (TRU) waste, spent nuclear fuel (SNF), and hazardous waste. Unit rates for transportation of contact-handled (<200 mrem/hr contact dose) and remote-handled (>200 mrem/hr contact dose) radioactive waste are estimated. Land transportation of radioactive and hazardous waste is subject to regulations promulgated by DOE, the U.S. Department of Transportation (DOT), the U.S. Nuclear Regulatory Commission (NRC), and state and local agencies. The cost estimates in this report assume compliance with applicable regulations.

  2. Comprehensive testing to measure the response of fluorocarbon rubber (FKM) to Hanford tank waste simulant

    SciTech Connect (OSTI)

    NIGREY,PAUL J.; BOLTON,DENNIS L.

    2000-02-01T23:59:59.000Z

    This report presents the findings of the Chemical Compatibility Program developed to evaluate plastic packaging components that may be incorporated in packaging mixed-waste forms for transportation. Consistent with the methodology outlined in this report, the authors performed the second phase of this experimental program to determine the effects of simulant Hanford tank mixed wastes on packaging seal materials. That effort involved the comprehensive testing of five plastic liner materials in an aqueous mixed-waste simulant. The testing protocol involved exposing the materials to {approximately}143, 286, 571, and 3,670 Krad of gamma radiation and was followed by 7-, 14-, 28-, 180-day exposures to the waste simulant at 18, 50, and 60 C. Fluorocarbon (FKM) rubber samples subjected to the same protocol were then evaluated by measuring seven material properties: specific gravity, dimensional changes, mass changes, hardness, compression set, vapor transport rates, and tensile properties. From the analyses, they determined that FKM rubber is not a good seal material to withstand aqueous mixed wastes having similar composition to the one used in this study. They have determined that FKM rubber has limited chemical durability after exposure to gamma radiation followed by exposure to the Hanford tank simulant mixed waste at elevated temperatures above 18 C.

  3. Effects of simulant mixed waste on EPDM and butyl rubber

    SciTech Connect (OSTI)

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

    1997-11-01T23:59:59.000Z

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

  4. RH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions, LLC

    2003-08-25T23:59:59.000Z

    The purpose of this program guidance document is to provide technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package 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 SARP and/or 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 10 CFR {section} 71.11, ''Deliberate Misconduct.'' 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. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions. Following these instructions assures that operations are safe and meet the requirements of the SARP. This document is available on the Internet at: ttp://www.ws/library/t2omi/t2omi.htm. Users are responsible for ensuring they are using the current revision and change notices. Sites may prepare their own document using the word-for-word steps in th is document, in sequence, including Notes and cautions. Site specific information may be included as necessary. The document, and revisions, must then be submitted to CBFO at sitedocuments@wipp.ws for approval. A copy of the approval letter from CBFO shall be available for audit purposes. Users may develop site-specific procedures addressing preoperational activities, quality assurance (QA), hoisting and rigging, and radiation health physics to be used with the instructions contained in this document. Users may recommend changes to this document by submitting their recommendations (in writing) to the WIPP M&O Contractor RH Packaging Maintenance Engineer for evaluation. If approved, the change(s) will be incorporated into this document for use by ALL users. Before first use and every 12 months after, user sites will be audited to this document to ensure compliance. They will also be audited within one year from the effective date of revisions to this document.

  5. Comparison of Different Upscaling Methods for Predicting Thermal Conductivity of Complex Heterogeneous Materials System: Application on Nuclear Waste Forms

    SciTech Connect (OSTI)

    Li, Dongsheng; Sun, Xin; Khaleel, Mohammad A.

    2012-06-16T23:59:59.000Z

    To develop a strategy in thermal conductivity prediction of a complex heterogeneous materials system, loaded nuclear waste forms, the computational efficiency and accuracy of different upscaling methods have been evaluated. The effective thermal conductivity, obtained from microstructure information and local thermal conductivity of different components, is critical in predicting the life and performance of waste form during storage. Several methods, including the Taylor model, Sachs model, self-consistent model, and statistical upscaling method, were developed and implemented. Microstructure based finite element method (FEM) prediction results were used to as benchmark to determine the accuracy of the different upscaling methods. Micrographs from waste forms with varying waste loadings were used in the prediction of thermal conductivity in FEM and homogenization methods. Prediction results demonstrated that in term of efficiency, boundary models (e.g., Taylor model and Sachs model) are stronger than the self-consistent model, statistical upscaling method, and finite element method. However, when balancing computational efficiency and accuracy, statistical upscaling is a useful method in predicting effective thermal conductivity for nuclear waste forms.

  6. Packaging of solid state devices

    DOE Patents [OSTI]

    Glidden, Steven C.; Sanders, Howard D.

    2006-01-03T23:59:59.000Z

    A package for one or more solid state devices in a single module that allows for operation at high voltage, high current, or both high voltage and high current. Low thermal resistance between the solid state devices and an exterior of the package and matched coefficient of thermal expansion between the solid state devices and the materials used in packaging enables high power operation. The solid state devices are soldered between two layers of ceramic with metal traces that interconnect the devices and external contacts. This approach provides a simple method for assembling and encapsulating high power solid state devices.

  7. WASTE POLICY STATEMENT 1 The University of Aberdeen's Waste Policy forms part of the institutional commitment to constant

    E-Print Network [OSTI]

    Levi, Ran

    supply chain; To encourage take-back schemes for WEEE, packaging and other wastes; To reuse and repair

  8. Testing of organic waste surrogate materials in support of the Hanford organic tank program. Final supplementary report, Supplement 1

    SciTech Connect (OSTI)

    Turner, D.A. [Westinghouse Hanford Co., Richland, WA (United States); Miron, Y. [Bureau of Mines (United States)

    1994-01-01T23:59:59.000Z

    To address safety issues regarding effective waste management efforts of underground organic waste storage tanks at the Hanford Site, the Bureau of Mines conducted a series of supplemental tests, at the request of the Westinghouse Hanford Company. In this series of supplemental tests, the thermal characteristics of surrogate materials, chosen by Hanford, were determined. The surrogate materials were mixtures of organic and inorganic sodium salts, representing fuels and oxidants. The oxidants were sodium nitrate and sodium nitrite. The fuels were sodium salts of acetate, stearate, and oxalate. Sodium hydroxide was also an ingredient of the mixtures, used to maintain basic conditions. The fuels and all the mixtures of this report were tested by differential scanning calorimetry and by thermogravimetric analysis.

  9. Microwave thawing package and method

    DOE Patents [OSTI]

    Fathi, Zakaryae; Lauf, Robert J.

    2004-03-16T23:59:59.000Z

    A package for containing frozen liquids during an electromagnetic thawing process includes: a first section adapted for containing a frozen material and exposing the frozen material to electromagnetic energy; a second section adapted for receiving thawed liquid material and shielding the thawed liquid material from further exposure to electromagnetic energy; and a fluid communication means for allowing fluid flow between the first section and the second section.

  10. Summary of radioactive solid waste received in the 200 Areas during calendar year 1994

    SciTech Connect (OSTI)

    Anderson, J.D.; Hagel, D.L.

    1995-08-01T23:59:59.000Z

    Westinghouse Hanford Company manages and operates the Hanford Site 200 Area radioactive solid waste storage and disposal facilities for the US Department of Energy, Richland Field Office, under contract DE-AC06-87RL10930. These facilities include radioactive solid waste disposal sites and radioactive solid waste storage areas. This document summarizes the amount of radioactive material that has been buried and stored in the 200 Area radioactive solid waste storage and disposal facilities from startup in 1944 through calendar year 1994. This report does not include backlog waste: solid radioactive wastes in storage or disposed of in other areas or facilities such as the underground tank farms. Unless packaged within the scope of WHC-EP-0063, Hanford Site Solid Waste Acceptance Criteria (WHC 1988), liquid waste data are not included in this document.

  11. Summary of radioactive solid waste received in the 200 Areas during calendar year 1993

    SciTech Connect (OSTI)

    Anderson, J.D.; Hagel, D.L.

    1994-09-01T23:59:59.000Z

    Westinghouse Hanford Company manages and operates the Hanford Site 200 Areas radioactive solid waste storage and disposal facilities for the US Department of Energy, Richland Operations Office. These facilities include radioactive solid waste disposal sites and radioactive solid waste storage areas. This document summarizes the amount of radioactive materials that have been buried and stored in the 200 Areas radioactive solid waste storage and disposal facilities since startup in 1944 through calendar year 1993. This report does not include backlog waste, solid radioactive waste in storage or disposed of in other areas, or facilities such as the underground tank farms. Unless packaged within the scope of WHC-EP-0063, ``Hanford Site Solid Waste Acceptance Criteria,`` (WHC 1988), liquid waste data are not included in this document.

  12. Urban Mining: Quality and quantity of recyclable and recoverable material mechanically and physically extractable from residual waste

    SciTech Connect (OSTI)

    Di Maria, Francesco, E-mail: francesco.dimaria@unipg.it; Micale, Caterina; Sordi, Alessio; Cirulli, Giuseppe; Marionni, Moreno

    2013-12-15T23:59:59.000Z

    Highlights: • Material recycling and recovery from residual waste by physical and mechanical process has been investigated. • About 6% of recyclable can be extracted by NIR and 2-3Dimension selector. • Another 2% of construction materials can be extracted by adopting modified soil washing process. • Extracted material quality is quite high even some residual heavy metal have been detected by leaching test. - Abstract: The mechanically sorted dry fraction (MSDF) and Fines (<20 mm) arising from the mechanical biological treatment of residual municipal solid waste (RMSW) contains respectively about 11% w/w each of recyclable and recoverable materials. Processing a large sample of MSDF in an existing full-scale mechanical sorting facility equipped with near infrared and 2-3 dimensional selectors led to the extraction of about 6% w/w of recyclables with respect to the RMSW weight. Maximum selection efficiency was achieved for metals, about 98% w/w, whereas it was lower for Waste Electrical and Electronic Equipment (WEEE), about 2% w/w. After a simulated lab scale soil washing treatment it was possible to extract about 2% w/w of inert exploitable substances recoverable as construction materials, with respect to the amount of RMSW. The passing curve showed that inert materials were mainly sand with a particle size ranging from 0.063 to 2 mm. Leaching tests showed quite low heavy metal concentrations with the exception of the particles retained by the 0.5 mm sieve. A minimum pollutant concentration was in the leachate from the 10 and 20 mm particle size fractions.

  13. Waste products in highway construction. Final report

    SciTech Connect (OSTI)

    Han, C.

    1993-04-01T23:59:59.000Z

    The report presents waste materials and products for highway construction. The general legislation, local liability, and research projects related to waste materials are outlined. The waste materials and products presented include waste paving materials, industrial ash materials, taconite tailing materials, waste tire rubber materials and products, building rubble materials, incinerator ash products and materials, waste glass materials, waste shingle materials and products, waste plastics products, and slag materials. For each waste category, the legislation and restrictions, material properties, construction and application, field performance, and recycling at the end of service life if available are discussed.

  14. Geochemical Processes Data Package for the Vadose Zone in the Single-Shell Tank Waste Management Areas at the Hanford Site

    SciTech Connect (OSTI)

    Cantrell, Kirk J.; Zachara, John M.; Dresel, P. Evan; Krupka, Kenneth M.; Serne, R. Jeffrey

    2007-09-28T23:59:59.000Z

    This data package discusses the geochemistry of vadose zone sediments beneath the single-shell tank 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 process information available for the vadose zone beneath the single-shell tank farms and the Integrated Disposal Facility. Two companion reports to this one were recently published which discuss the geology of the farms (Reidel and Chamness 2007) and groundwater flow and contamination beneath the farms (Horton 2007).

  15. Co-gasification of municipal solid waste and material recovery in a large-scale gasification and melting system

    SciTech Connect (OSTI)

    Tanigaki, Nobuhiro, E-mail: tanigaki.nobuhiro@nsc-eng.co.jp [Nippon Steel Engineering Co., Ltd. (Head Office), Osaki Center Building 1-5-1, Osaki, Shinagawa-ku, Tokyo 141-8604 (Japan); Manako, Kazutaka [Nippon Steel Engineering Co., Ltd., 46-59, Nakabaru, Tobata-ku, Kitakyushu, Fukuoka 804-8505 (Japan); Osada, Morihiro [Nippon Steel Engineering Co., Ltd. (Head Office), Osaki Center Building 1-5-1, Osaki, Shinagawa-ku, Tokyo 141-8604 (Japan)

    2012-04-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer This study evaluates the effects of co-gasification of MSW with MSW bottom ash. Black-Right-Pointing-Pointer No significant difference between MSW treatment with and without MSW bottom ash. Black-Right-Pointing-Pointer PCDD/DFs yields are significantly low because of the high carbon conversion ratio. Black-Right-Pointing-Pointer Slag quality is significantly stable and slag contains few hazardous heavy metals. Black-Right-Pointing-Pointer The final landfill amount is reduced and materials are recovered by DMS process. - Abstract: This study evaluates the effects of co-gasification of municipal solid waste with and without the municipal solid waste bottom ash using two large-scale commercial operation plants. From the viewpoint of operation data, there is no significant difference between municipal solid waste treatment with and without the bottom ash. The carbon conversion ratios are as high as 91.7% and 95.3%, respectively and this leads to significantly low PCDD/DFs yields via complete syngas combustion. The gross power generation efficiencies are 18.9% with the bottom ash and 23.0% without municipal solid waste bottom ash, respectively. The effects of the equivalence ratio are also evaluated. With the equivalence ratio increasing, carbon monoxide concentration is decreased, and carbon dioxide and the syngas temperature (top gas temperature) are increased. The carbon conversion ratio is also increased. These tendencies are seen in both modes. Co-gasification using the gasification and melting system (Direct Melting System) has a possibility to recover materials effectively. More than 90% of chlorine is distributed in fly ash. Low-boiling-point heavy metals, such as lead and zinc, are distributed in fly ash at rates of 95.2% and 92.0%, respectively. Most of high-boiling-point heavy metals, such as iron and copper, are distributed in metal. It is also clarified that slag is stable and contains few harmful heavy metals such as lead. Compared with the conventional waste management framework, 85% of the final landfill amount reduction is achieved by co-gasification of municipal solid waste with bottom ash and incombustible residues. These results indicate that the combined production of slag with co-gasification of municipal solid waste with the bottom ash constitutes an ideal approach to environmental conservation and resource recycling.

  16. Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal

    SciTech Connect (OSTI)

    Not Available

    1984-07-01T23:59:59.000Z

    The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

  17. Solid waste handling

    SciTech Connect (OSTI)

    Parazin, R.J.

    1995-05-31T23:59:59.000Z

    This study presents estimates of the solid radioactive waste quantities that will be generated in the Separations, Low-Level Waste Vitrification and High-Level Waste Vitrification facilities, collectively called the Tank Waste Remediation System Treatment Complex, over the life of these facilities. This study then considers previous estimates from other 200 Area generators and compares alternative methods of handling (segregation, packaging, assaying, shipping, etc.).

  18. Comprehensive testing to measure the response of butyl rubber to Hanford tank waste simulant

    SciTech Connect (OSTI)

    NIGREY,PAUL J.

    2000-05-01T23:59:59.000Z

    This report presents the findings of the Chemical Compatibility Program developed to evaluate plastic packaging components that may be incorporated in packaging mixed-waste forms for transportation. Consistent with the methodology outlined in this report, the authors performed the second phase of this experimental program to determine the effects of simulant Hanford tank mixed wastes on packaging seal materials. That effort involved the comprehensive testing of five plastic liner materials in an aqueous mixed-waste simulant. The testing protocol involved exposing the materials to {approximately}143, 286, 571, and 3,670 krad of gamma radiation and was followed by 7-, 14-, 28-, 180-day exposures to the waste simulant at 18, 50, and 60 C. Butyl rubber samples subjected to the same protocol were then evaluated by measuring seven material properties: specific gravity, dimensional changes, mass changes, hardness, compression set, vapor transport rates, and tensile properties. From the analyses, they determined that butyl rubber has relatively good resistance to radiation, this simulant, and a combination of these factors. These results suggest that butyl rubber is a relatively good seal material to withstand aqueous mixed wastes having similar composition to the one used in this study.

  19. Response of ethylene propylene diene monomer rubber (EPDM) to simulant Hanford tank waste

    SciTech Connect (OSTI)

    NIGREY,PAUL J.

    2000-02-01T23:59:59.000Z

    This report presents the findings of the Chemical Compatibility Program developed to evaluate plastic packaging components that may be incorporated in packaging mixed-waste forms for transportation. Consistent with the methodology outlined in this report, the author performed the second phase of this experimental program to determine the effects of simulant Hanford tank mixed wastes on packaging seal materials. That effort involved the comprehensive testing of five plastic liner materials in an aqueous mixed-waste simulant. The testing protocol involved exposing the materials to {approximately}143, 286, 571, and 3,670 krad of gamma radiation and was followed by 7-, 14-, 28-, 180-day exposures to the waste simulant at 18, 50, and 60 C. Ethylene propylene diene monomer (EPDM) rubber samples subjected to the same protocol were then evaluated by measuring seven material properties: specific gravity, dimensional changes, mass changes, hardness, compression set, vapor transport rates, and tensile properties. The author has determined that EPDM rubber has excellent resistance to radiation, this simulant, and a combination of these factors. These results suggest that EPDM is an excellent seal material to withstand aqueous mixed wastes having similar composition to the one used in this study.

  20. A data base and a standard material for use in acceptance testing of low-activity waste products

    SciTech Connect (OSTI)

    Wolf, S.F.; Ebert, W.L.; Luo, J.S.; Strachan, D.M.

    1998-04-01T23:59:59.000Z

    The authors have conducted replicate dissolution tests following the product consistency test (PCT) procedure to measure the mean and standard deviation of the solution concentrations of B, Na, and Si at various combinations of temperature, duration, and glass/water mass ratio. Tests were conducted with a glass formulated to be compositionally similar to low-activity waste products anticipated for Hanford to evaluate the adequacy of test methods that have been designated in privatization contracts for use in product acceptance. An important finding from this set of tests is that the solution concentrations generated in tests at 20 C will likely be too low to measure the dissolution rates of waste products reliably. Based on these results, the authors recommend that the acceptance test be conducted at 40 C. Tests at 40 C generated higher solution concentrations, were more easily conducted, and the measured rates were easily related to those at 20 C. Replicate measurements of other glass properties were made to evaluate the possible use of LRM-1 as a standard material. These include its composition, homogeneity, density, compressive strength, the Na leachability index with the ANSI/ANS 16.1 leach test, and if the glass is characteristically hazardous with the toxicity characteristic leach procedure. The values of these properties were within the acceptable limits identified for Hanford low-activity waste products. The reproducibility of replicate tests and analyses indicates that the glass would be a suitable standard material.

  1. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers; Overview

    SciTech Connect (OSTI)

    Farmer, J.C.; McCright, R.D.; Kass, J.N.

    1988-06-01T23:59:59.000Z

    Three iron- to nickel-based austenitic alloys and three copper-based alloys are being considered as candidate materials for the fabrication of high-level radioactive-waste disposal containers. The austenitic alloys are Types 304L and 316L stainless steels and the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). Waste in the forms of both spent fuel assemblies from reactors and borosilicate glass will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking; and transgranular stress corrosion cracking. Problems specific to welds, such as hot cracking, may also occur. A survey of the literature has been prepared as part of the process of selecting, from among the candidates, a material that is adequate for repository conditions. The modes of degradation are discussed in detail in the survey to determine which apply to the candidate alloys and the extent to which they may actually occur. The eight volumes of the survey are summarized in Sections 1 through 8 of this overview. The conclusions drawn from the survey are also given in this overview.

  2. Remaining Sites Verification Package for the 100-F-26:13, 108-F Drain Pipelines, Waste Site Reclassification Form 2005-011

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-03-03T23:59:59.000Z

    The 100-F-26:13 waste site is the network of process sewer pipelines that received effluent from the 108-F Biological Laboratory and discharged it to the 188-F Ash Disposal Area (126-F-1 waste site). The pipelines included one 0.15-m (6-in.)-, two 0.2-m (8-in.)-, and one 0.31-m (12-in.)-diameter vitrified clay pipe segments encased in concrete. 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.

  3. Remaining Sites Verification Package for the 1607-B2 Septic System and 100-B-14:2 Sanitary Sewer System, Waste Site Reclassification Form 2006-055

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-03-21T23:59:59.000Z

    The 1607-B2 waste site is a former septic system associated with various 100-B facilities, including the 105-B, 108-B, 115-B/C, and 185/190-B buildings. The site was evaluated based on confirmatory results for feeder lines within the 100-B-14:2 subsite and determined to require remediation. The 1607-B2 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.

  4. RADIONUCLIDE DATA PACKAGE FOR PERFORMANCE ASSESSMENT CALCULATIONS RELATED TO THE E-AREA LOW-LEVEL WASTE FACILITY AT THE SAVANNAH RIVER SITE.

    SciTech Connect (OSTI)

    Cook, J

    2007-03-20T23:59:59.000Z

    The Savannah River Site disposes of low-level radioactive waste within on-site engineered disposal facilities. The Savannah River Site must demonstrate that these disposals meet the requirements of DOE Order 435 . 1 through a process known as performance assessment (PA). The objective of this document is to provide the radionuclide -specific data needed for the PA calculations . This work is part of an on-going program to periodically review and update existing PA work as new data becomes available. Revision of the E -Area Low-Level Waste Facility PA is currently underway. The number of radionuclides selected to undergo detailed analysis in the PA is determined by a screening process. The basis of this process is described. Radionuclide-specific data for half-lives, decay modes, daughters, dose conversion factors and groundwater concentration limits are presented with source references and methodologies.

  5. Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway, Waste Site Reclassification Form 2006-038

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-09-25T23:59:59.000Z

    The 116-F-8 waste site is the former 1904-F Outfall Structure used to discharge reactor cooling water effluent fro mthe 107-F Retention Basin to the Columbia River. 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.

  6. Order Module--DOE O 460.1C, PACKAGING AND TRANSPORTATION SAFETY...

    Office of Environmental Management (EM)

    DOE O 460.1C, PACKAGING AND TRANSPORTATION SAFETY, DOE O 460.2A, DEPARTMENTAL MATERIALS TRANSPORTATION AND PACKAGING MANAGEMENT Order Module--DOE O 460.1C, PACKAGING AND...

  7. An assessment of the value of retail ready packaging

    E-Print Network [OSTI]

    Jackson, Kathleen Anne

    2008-01-01T23:59:59.000Z

    Use of retail-ready packaging reduces the costs of replenishing store shelves by eliminating the labor of removing packaging materials and stocking individual items on shelves. While reducing costs for retailers, retail-ready ...

  8. Safety analysis report for packaging (onsite) steel drum

    SciTech Connect (OSTI)

    McCormick, W.A.

    1998-09-29T23:59:59.000Z

    This Safety Analysis Report for Packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the steel drum packaging system meets the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments, for an onsite packaging containing Type B quantities of solid and liquid radioactive materials. The basic component of the steel drum packaging system is the 208 L (55-gal) steel drum.

  9. Remaining Sites Verification Package for the 126-B-3, 184-B Coal Pit Dumping Area, Waste Site Reclassification Form 2005-028

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-08-07T23:59:59.000Z

    The 126-B-3 waste site is the former coal storage pit for the 184-B Powerhouse. During demolition operations in the 1970s, the site was used for disposal of demolition debris from 100-B/C Area facilities. The site has been remediated by removing debris and contaminated soils. 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.

  10. Remaining Sites Verification Package for the 600-243 Petroleum-Contaminated Soil Bioremediation Pad, Waste Site Reclassification Form 2007-033

    SciTech Connect (OSTI)

    J. M. Capron

    2008-11-07T23:59:59.000Z

    The 600-243 waste site consisted of a bioremediation pad for petroleum-contaminated soils resulting from the 1100 Area Underground Storage Tank (UST) upgrades in 1994. 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. Remaining Sites Verification Package for the 100-F-33, 146-F Aquatic Biology Fish Ponds, Waste Site Reclassification Form 2006-021

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-08-25T23:59:59.000Z

    The 100-F-33, 146-F Aquatice Biology Fish Ponds waste site was an area with six small rectangular ponds and one large circular pond used to conduct tests on fish using various mixtures of river and reactor effluent water. The current site conditions achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification and applicable 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.

  12. Remaining Sites Verification Package for 132-H-1, 116-H Reactor Stack Burial Site, Waste Site Reclassification Form 2006-053

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-06-26T23:59:59.000Z

    The 132-H-1 waste site includes the 116-H exhaust stack burial trench and the buried stack foundation (which contains an embedded vertical 15-cm (6-in) condensate drain line). The 116-H reactor exhaust stack and foundation were decommissioned and demolished using explosives in 1983, with the rubble buried in situ beneath clean fill at least 1 m (3.3 ft) thick. 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.

  13. Remaining Sites Verification Package for the 1607-F7, 141-M Building Septic Tank, Waste Site Reclassification Form 2006-040

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-10-19T23:59:59.000Z

    The 1607-F7, 141-M Building Septic Tank waste site was a septic tank and drain field that received sanitary sewage from the former 141-M Building. Remedial action was performed in August and November 2005. The results of verification sampling demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. These results also show that residual concentrations support unrestricted future use of shallow zone soil and that contaminant levels remaining in the soil are protective of groundwater and the Columbia River.

  14. THE SUCCESSFUL UTILIZATION OF COMMERCIAL TREATMENT CAPABILITIES TO DISPOSITION HANFORD NO-PATH-FORWARD SUSPECT TRANSURANIC WASTES

    SciTech Connect (OSTI)

    BLACKFORD LT; CATLOW RL; WEST LD; COLLINS MS; ROMINE LD; MOAK DJ

    2012-01-30T23:59:59.000Z

    The U.S. Department of Energy (DOE) Richland Operations Office (RL) has adopted the 2015 Vision for Cleanup of the Hanford Site. The CH2M HILL Plateau Remediation Company's (CHPRC) Waste and Fuels Management Project (W&FMP) and their partners support this mission by providing centralized waste management services for the Hanford Site waste generating organizations. At the time of the CHPRC contract award (August 2008) slightly more than 9,000 cubic meters (m{sup 3}) of legacy waste was defined as ''no-path-forward waste.'' A significant portion of this waste (7,650 m{sup 3}) comprised wastes with up to 50 grams of special nuclear materials (SNM) in oversized packages recovered during retrieval operations and large glove boxes removed from Hanford's Plutonium Finishing Plant (PFP). Through a collaborative effort between the DOE, CHPRC, and Perma-Fix Environmental Services, Inc. (PESI), pathways for these problematic wastes were developed and are currently being implemented.

  15. Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

    SciTech Connect (OSTI)

    Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D. [Australian Nuclear Science and Technology Organisation (ANSTO), New Illawarra Road, Lucas Heights, NSW 2234 (Australia)] [Australian Nuclear Science and Technology Organisation (ANSTO), New Illawarra Road, Lucas Heights, NSW 2234 (Australia); Scales, Charlie R.; Maddrell, Ewan R. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom)] [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom); Hobbs, Jeff [Sellafield Limited, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom)] [Sellafield Limited, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom)

    2013-07-01T23:59:59.000Z

    A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

  16. Oxidation/corrosion of metallic and ceramic materials in an aluminum remelt furnace. [For fluidized bed waste heat recovery systems

    SciTech Connect (OSTI)

    Federer, J.I.; Jones, P.J.

    1985-12-01T23:59:59.000Z

    Both metallic alloys and ceramic materials are candidates for the distributor plate and other components of fluidized bed waste heat recovery (FBWHR) systems. Eleven Fe-, Ni-, and Co-base alloys were exposed to air at elevated temperatures in laboratory furnaces and to flue gases in an aluminum remelt furnace to assess their resistance to oxidation and corrosion. Four SiC ceramics and two oxide ceramics were also tested in the aluminum remelt furnace. Some alloys were coated with aluminum or SiO2 by commercial processes in an effort to enhance their oxidation and corrosion resistance.

  17. Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings

    SciTech Connect (OSTI)

    DiSabatino, A; Biswas, D; DeMicco, M; Fisher, L E; Hafner, R; Haslam, J; Mok, G; Patel, C; Russell, E

    2007-04-12T23:59:59.000Z

    This Packaging Review Guide (PRG) provides guidance for Department of Energy (DOE) review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE Order 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his or her review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. This PRG is generally organized at the section level in a format similar to that recommended in Regulatory Guide 7.9 (RG 7.9). One notable exception is the addition of Section 9 (Quality Assurance), which is not included as a separate chapter in RG 7.9. Within each section, this PRG addresses the technical and regulatory bases for the review, the manner in which the review is accomplished, and findings that are generally applicable for a package that meets the approval standards. This Packaging Review Guide (PRG) provides guidance for DOE review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE O 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. The primary objectives of this PRG are to: (1) Summarize the regulatory requirements for package approval; (2) Describe the technical review procedures by which DOE determines that these requirements have been satisfied; (3) Establish and maintain the quality and uniformity of reviews; (4) Define the base from which to evaluate proposed changes in scope and requirements of reviews; and (5) Provide the above information to DOE organizations, contractors, other government agencies, and interested members of the general public. This PRG was originally published in September 1987. Revision 1, issued in October 1988, added new review sections on quality assurance and penetrations through the containment boundary, along with a few other items. Revision 2 was published October 1999. Revision 3 of this PRG is a complete update, and supersedes Revision 2 in its entirety.

  18. Glass Ceramic Formulation Data Package

    SciTech Connect (OSTI)

    Crum, Jarrod V.; Rodriguez, Carmen P.; McCloy, John S.; Vienna, John D.; Chung, Chul-Woo

    2012-06-17T23:59:59.000Z

    A glass ceramic waste form is being developed for treatment of secondary waste streams generated by aqueous reprocessing of commercial used nuclear fuel (Crum et al. 2012b). The waste stream contains a mixture of transition metals, alkali, alkaline earths, and lanthanides, several of which exceed the solubility limits of a single phase borosilicate glass (Crum et al. 2009; Caurant et al. 2007). A multi-phase glass ceramic waste form allows incorporation of insoluble components of the waste by designed crystallization into durable heat tolerant phases. The glass ceramic formulation and processing targets the formation of the following three stable crystalline phases: (1) powellite (XMoO4) where X can be (Ca, Sr, Ba, and/or Ln), (2) oxyapatite Yx,Z(10-x)Si6O26 where Y is alkaline earth, Z is Ln, and (3) lanthanide borosilicate (Ln5BSi2O13). These three phases incorporate the waste components that are above the solubility limit of a single-phase borosilicate glass. The glass ceramic is designed to be a single phase melt, just like a borosilicate glass, and then crystallize upon slow cooling to form the targeted phases. The slow cooling schedule is based on the centerline cooling profile of a 2 foot diameter canister such as the Hanford High-Level Waste canister. Up to this point, crucible testing has been used for glass ceramic development, with cold crucible induction melter (CCIM) targeted as the ultimate processing technology for the waste form. Idaho National Laboratory (INL) will conduct a scaled CCIM test in FY2012 with a glass ceramic to demonstrate the processing behavior. This Data Package documents the laboratory studies of the glass ceramic composition to support the CCIM test. Pacific Northwest National Laboratory (PNNL) measured melt viscosity, electrical conductivity, and crystallization behavior upon cooling to identify a processing window (temperature range) for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form.

  19. Research and Development of a New Silica-Alumina Based Cementitious Material Largely Using Coal Refuse for Mine Backfill, Mine Sealing and Waste Disposal Stabilization

    SciTech Connect (OSTI)

    Henghu Sun; Yuan Yao

    2012-06-29T23:59:59.000Z

    Coal refuse and coal combustion byproducts as industrial solid waste stockpiles have become great threats to the environment. To activate coal refuse is one practical solution to recycle this huge amount of solid waste as substitute for Ordinary Portland Cement (OPC). The central goal of this project is to investigate and develop a new silica-alumina based cementitious material largely using coal refuse as a constituent that will be ideal for durable construction, mine backfill, mine sealing and waste disposal stabilization applications. This new material is an environment-friendly alternative to Ordinary Portland Cement. The main constituents of the new material are coal refuse and other coal wastes including coal sludge and coal combustion products (CCPs). Compared with conventional cement production, successful development of this new technology could potentially save energy and reduce greenhouse gas emissions, recycle vast amount of coal wastes, and significantly reduce production cost. A systematic research has been conducted to seek for an optimal solution for enhancing pozzolanic reactivity of the relatively inert solid waste-coal refuse in order to improve the utilization efficiency and economic benefit as a construction and building material.

  20. Modeling Coupled Processes in Clay Formations for Radioactive Waste Disposal

    E-Print Network [OSTI]

    Liu, Hui-Hai

    2010-01-01T23:59:59.000Z

    sorption including waste heat, hyperalkaline solutions frome.g. , heat production from the decay of the waste, re-waste packages along the tunnels, to achieve a distributed heat

  1. Performance oriented packaging testing of nine Mk 3 Mod 0 signal containers in PPP-B-621 wood box for packing group II solid hazardous materials. Final report

    SciTech Connect (OSTI)

    Libbert, K.J.

    1992-10-01T23:59:59.000Z

    A PPP-B-621 wood box containing nine Mk 3 Mod 0 Signal containers was tested for conformance to Performance Oriented Packaging criteria established by Code of Federal Regulations Title 49 CFR. The container was tested with a gross weight of 123.3 pounds (56 kilograms) and met all requirements.

  2. ETC/RWM working paper 2005/1 Outlook for waste and material

    E-Print Network [OSTI]

    ...............................................................................22 3.3. Model for fossil fuel material flows.......................................................................................................84 5.3. DMC ­ Fossil fuels

  3. Remaining Sites Verification Package for the 100-C-9:1 Main Process Sewer Collection Line, Waste Site Reclassification Form 2004-012

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-06-11T23:59:59.000Z

    The 100-C-9:1 main process sewer pipeline, also known as the twin box culvert, was a dual reinforced process sewer that collected process effluent from the 183-C and 190-C water treatment facilities, discharging at the 132-C-2 Outfall. For remedial action purposes, the 100-C-9:1 waste site was subdivided into northern and southern sections. The 100-C-9:1 subsite 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.

  4. Remaining Sites Verification Package for the 100-F-44:2, Discovery Pipeline Near 108-F Building, Waste Site Reclassification Form 2007-006

    SciTech Connect (OSTI)

    J. M. Capron

    2008-05-30T23:59:59.000Z

    The 100-F-44:2 waste site is a steel pipeline that was discovered in a junction box during confirmatory sampling of the 100-F-26:4 pipeline from December 2004 through January 2005. The 100-F-44:2 pipeline feeds into the 100-F-26:4 subsite vitrified clay pipe (VCP) process sewer pipeline from the 108-F Biology Laboratory at the junction box. 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.

  5. Remaining Sites Verification Package for the 100-F-26:12, 1.8-m (72-in.) Main Process Sewer Pipeline, Waste Site Reclassification Form 2007-034

    SciTech Connect (OSTI)

    J. M. Capron

    2008-04-29T23:59:59.000Z

    The 100-F-26:12 waste site was an approximately 308-m-long, 1.8-m-diameter east-west-trending reinforced concrete pipe that joined the North Process Sewer Pipelines (100-F-26:1) and the South Process Pipelines (100-F-26:4) with the 1.8-m reactor cooling water effluent pipeline (100-F-19). 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.

  6. Cost Estimation Package

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1997-03-28T23:59:59.000Z

    This chapter focuses on the components (or elements) of the cost estimation package and their documentation.

  7. Cleanup Verification Package for the118-F-2 Burial Ground

    SciTech Connect (OSTI)

    J. M. Capron and K. A. Anselm

    2008-02-21T23:59:59.000Z

    This cleanup verification package documents completion of remedial action, sampling activities, and compliance with cleanup criteria for the 118-F-2 Burial Ground. This burial ground, formerly called Solid Waste Burial Ground No. 1, was the original solid waste disposal site for the 100-F Area. Eight trenches contained miscellaneous solid waste from the 105-F Reactor and one trench contained solid waste from the biology facilities.

  8. Challenges in the Packaging of MEMS

    SciTech Connect (OSTI)

    BROWN, WILLIAM D.; EATON, WILLIAM P.; MALSHE, AJAY P.; MILLER, WILLIAM M.; O'NEAL, CHAD; SINGH, SUSHILA B.

    1999-09-24T23:59:59.000Z

    Microelectromechanical Systems (MEMS) packaging is much different from conventional integrated circuit (IC) packaging. Many MEMS devices must interface to the environment in order to perform their intended function, and the package must be able to facilitate access with the environment while protecting the device. The package must also not interfere with or impede the operation of the MEMS device. The die attachment material should be low stress, and low outgassing, while also minimizing stress relaxation overtime which can lead to scale factor shifts in sensor devices. The fabrication processes used in creating the devices must be compatible with each other, and not result in damage to the devices. Many devices are application specific requiring custom packages that are not commercially available. Devices may also need media compatible packages that can protect the devices from harsh environments in which the MEMS device may operate. Techniques are being developed to handle, process, and package the devices such that high yields of functional packaged parts will result. Currently, many of the processing steps are potentially harmful to MEMS devices and negatively affect yield. It is the objective of this paper to review and discuss packaging challenges that exist for MEMS systems and to expose these issues to new audiences from the integrated circuit packaging community.

  9. Relevance of biotic pathways to the long-term regulation of nuclear waste disposal. Estimation of radiation dose to man resulting from biotic transport: the BIOPORT/MAXI1 software package. Volume 5

    SciTech Connect (OSTI)

    McKenzie, D.H.; Cadwell, L.L.; Gano, K.A.; Kennedy, W.E. Jr.; Napier, B.A.; Peloquin, R.A.; Prohammer, L.A.; Simmons, M.A.

    1985-10-01T23:59:59.000Z

    BIOPORT/MAXI1 is a collection of five computer codes designed to estimate the potential magnitude of the radiation dose to man resulting from biotic transport processes. Dose to man is calculated for ingestion of agricultural crops grown in contaminated soil, inhalation of resuspended radionuclides, and direct exposure to penetrating radiation resulting from the radionuclide concentrations established in the available soil surface by the biotic transport model. This document is designed as both an instructional and reference document for the BIOPORT/MAXI1 computer software package and has been written for two major audiences. The first audience includes persons concerned with the mathematical models of biological transport of commercial low-level radioactive wastes and the computer algorithms used to implement those models. The second audience includes persons concerned with exercising the computer program and exposure scenarios to obtain results for specific applications. The report contains sections describing the mathematical models, user operation of the computer programs, and program structure. Input and output for five sample problems are included. In addition, listings of the computer programs, data libraries, and dose conversion factors are provided in appendices.

  10. RADIOACTIVE WASTE STORAGE IN MINED CAVERNS IN CRYSTALLINE ROCK-RESULTS OF FIELD INVESTIGATIONS AT STRIPA, SWEDEN

    E-Print Network [OSTI]

    Witherspoon, P.A.

    2010-01-01T23:59:59.000Z

    for Nuclear Waste Management, Materials Research Society.for Nuclear Waste Management, Materials Research Society.

  11. Exergy analysis of the Chartherm process for energy valorization and material recuperation of chromated copper arsenate (CCA) treated wood waste

    SciTech Connect (OSTI)

    Bosmans, A., E-mail: anouk.bosmans@mech.kuleuven.be [Department of Mechanical Engineering, Katholieke Universiteit Leuven, Celestijnenlaan 300A, 3001 Heverlee (Belgium); Auweele, M. Vanden; Govaerts, J.; Helsen, L. [Department of Mechanical Engineering, Katholieke Universiteit Leuven, Celestijnenlaan 300A, 3001 Heverlee (Belgium)

    2011-04-15T23:59:59.000Z

    The Chartherm process (Thermya, Bordeaux, France) is a thermochemical conversion process to treat chromated copper arsenate (CCA) impregnated wood waste. The process aims at maximum energy valorization and material recuperation by combining the principles of low-temperature slow pyrolysis and distillation in a smart way. The main objective of the exergy analysis presented in this paper is to find the critical points in the Chartherm process where it is necessary to apply some measures in order to reduce exergy consumption and to make energy use more economic and efficient. It is found that the process efficiency can be increased with 2.3-4.2% by using the heat lost by the reactor, implementing a combined heat and power (CHP) system, or recuperating the waste heat from the exhaust gases to preheat the product gas. Furthermore, a comparison between the exergetic performances of a 'chartherisation' reactor and an idealized gasification reactor shows that both reactors destroy about the same amount of exergy (i.e. 3500 kW kg{sub wood}{sup -1}) during thermochemical conversion of CCA-treated wood. However, the Chartherm process possesses additional capabilities with respect to arsenic and tar treatment, as well as the extra benefit of recuperating materials.

  12. Nuclear waste management. Semiannual progress report, October 1983-March 1984

    SciTech Connect (OSTI)

    McElroy, J.L.; Powell, J.A.

    1984-06-01T23:59:59.000Z

    Progress in the following studies on radioactive waste management is reported: defense waste technology; Nuclear Waste Materials Characterization Center; waste isolation; and supporting studies. 58 figures, 22 tables.

  13. The potential consequences and risks of highway accidents involving gamma-emitting low specific activity (LSA) waste

    SciTech Connect (OSTI)

    Ostmeyer, R.M.; Finley, N.C.; Cashwell, J.W.; McClure, J.D.

    1988-08-01T23:59:59.000Z

    This report documents an analysis of the potential risks and consequences of highway accidents that lead to the release of gamma-emitting low specific activity (LSA) materials from their packagings. Of particular concern is LSA waste with a specific activity level near the existing regulatory limit. The results of the analysis are intended primarily to provide a basis for evaluating limits on the quantity of radionuclides that can be contained in a single LSA waste shipment. 81 refs., 39 figs., 52 tabs.

  14. Y-12 defense programs: Nuclear Packaging Systems testing capabilities

    SciTech Connect (OSTI)

    NONE

    1995-06-01T23:59:59.000Z

    The Nuclear Packaging Systems (NPS) Department can manage/accomplish any packaging task. The NPS organization is responsible for managing the design, testing, certification, procurement, operation, refurbishment, maintenance, and disposal of packaging used to transport radioactive materials, other hazardous materials, and general cargoes on public roads and within the Oak Ridge Y-12 Plant. Additionally, the NPS Department has developed a Quality Assurance plan for all packaging, design and procurement of nonweapon shipping containers for radioactive materials, and design and procurement of performance-oriented packaging for hazardous materials. Further, the NPS Department is responsible for preparation and submittal of Safety Analysis Reports for Packaging (SARP). The NPS Department coordinates shipping container procurement and safety certification activities that have lead-times of up to two years. A Packaging Testing Capabilities Table at the Oak Ridge complex is included as a table.

  15. It Just Keeps Getting Better-Tru Waste Inventory

    SciTech Connect (OSTI)

    Lott, S.; Crawford, B.; McInroy, W.; Van Soest, G.; McTaggart, J.; Guerin, D. [Los Alamos National Laboratory-Carlsbad Operations, Carlsbad, NM (United States); Patterson, R. [U.S. Department of Energy Carlsbad, Field Office, Carlsbad, NM (United States)

    2008-07-01T23:59:59.000Z

    The Waste Isolation Pilot Plant (WIPP) opened on March 26, 1999, becoming the nation's first deep geologic repository for the permanent disposal of defense-generated transuranic (TRU) waste. In May 1998, the U. S. Environmental Protection Agency (EPA) certified WIPP and re-certified WIPP in March 2006. The knowledge of TRU waste inventory is fundamental to packaging, transportation, disposal strategies, resource allocation, and is also imperative when working in a regulatory framework. TRU waste inventory data are used to define the waste that will fill the WIPP repository in terms of volume, radionuclides, waste material parameters, other chemical components, and to model the impact of the waste on the performance of the WIPP over a 10,000-year evolution. The data that pertain to TRU waste is defined in the WIPP Land Withdrawal Act (LWA), as '..waste containing more that 100 nanocuries of alpha-emitting transuranic isotopes per gram of waste, with half-lives greater than 20 years..' Defining TRU waste further, the wastes are classified as either contact-handled (CH) or remote-handled (RH) TRU waste, depending on the dose rate at the surface of the waste container. CH TRU wastes are packaged with an external surface dose rate not greater than 200 milli-rem (mrem) per hour, while RH TRU wastes are packaged with an external surface dose rate of 200 mrem per hour or greater. The Los Alamos National Laboratory-Carlsbad Operations (LANL-CO) Inventory Team has developed a powerful new database, the Comprehensive Inventory Database (CID), to maintain the TRU waste inventory information. The CID is intended to replace the Transuranic Waste Baseline Inventory Database (TWBID), Revision 2.1, as the central inventory information repository for tracking all existing and potential (TRU) waste generated across the Department of Energy (DOE) TRU waste complex. It is also the source for information submitted for the Annual TRU Waste Inventory Reports some of which will be used in future Compliance Re-certification Applications (CRAs) for the WIPP. Currently, the DOE is preparing for the second re-certification, CRA-2009. The CID contains comprehensive TRU waste inventory that is consistent, relevant, and easily accessible to support DOE needs, not only the CRAs and performance assessments, but also waste management planning activities and other regulatory needs (e.g., National Environmental Policy Act (NEPA) analyses). The comprehensive inventory contains information obtained via inventory updates and approved acceptable knowledge (AK) characterization information to ensure inventory data integrity is maintained and the inventory is current. The TRU waste inventory is maintained in the CID under configuration management as defined in the LANL-CO Quality Assurance Program. The CID was developed using Microsoft{sup TM} Access Data Project{sup TM} (ADP) technology with a Microsoft SQL Server{sup TM} back end. The CID is user friendly, contains more fields, provides for easy upload of data, and has the capability to generate fully qualified data reports. To go along with the new database, the LANL-CO Inventory Team has developed an improved data collection/screening process and has excellent communications with the TRU waste site personnel. WIPP has now received over 6,000 shipments, emplaced over 50,000 cubic meters of CH waste, and successfully completed one re-certification. With a new robust qualified database, the CID, to maintain the inventory information, the TRU waste inventory information is continuously improving in quality, accuracy, and usability (better). (authors)

  16. Materials Characterization Paper In Support of the Proposed Rulemaking: Identification of Nonhazardous Secondary Materials That Are Solid Waste – Used Oil

    E-Print Network [OSTI]

    unknown authors

    2010-01-01T23:59:59.000Z

    EPA defines used oil as any oil that has been refined from crude oil, or any synthetic oil, that has been used and as a result of such use is contaminated by physical or chemical impurities. 1 EPA’s criteria for used oil: • Origin: Used oil must have been refined from crude oil or made from synthetic materials (i.e., derived from coal, shale, or polymers). Examples of crude-oil derived oils and synthetic oils are motor oil, mineral oil, laminating surface agents, and metal working oils. Thus, animal and vegetable oils are not included. Bottom clean-out from virgin fuel oil storage tanks or virgin oil recovered from a spill, as well as products solely used as cleaning agents or for their solvent properties, and certain petroleum-derived products such as antifreeze and kerosene are also not included. Use: The oil must have been used as a lubricant, coolant, heat (non-contact) transfer fluid, hydraulic fluid, heat transfer fluid or for a similar use. Lubricants include, but are not limited to, used motor oil, metal working lubricants, and emulsions. An example of a hydraulic fluid is transmission fluid. Heat transfer fluids can be materials such as coolants, heating media, refrigeration oils, and electrical insulation oils. Authorized states or regions determine what is considered a “similar use ” on a site-specific basis according to whether the material is used and managed in a manner consistent with Part 279 (e.g., used as a buoyant). Contaminants: The used oil must be contaminated by physical (e.g., high water content, metal shavings, or dirt) or chemical (e.g., lead, halogens, solvents or other hazardous constituents) impurities as a result of use. 2. Annual Quantities of Used Oil Generated and Used

  17. Borehole Data Package for RCRA Well 299-W22-47 at Single-Shell Tank Waste Management Area S-SX, Hanford Site, Washington

    SciTech Connect (OSTI)

    Horton, Duane G.; Chamness, Mickie A.

    2006-04-17T23:59:59.000Z

    One new Resource Conservation and Recovery Act (RCRA) groundwater assessment well was installed at single-shell tank Waste Management Area (WMA) S-SX in fiscal year (FY) 2005 to fulfill commitments for well installations proposed in Hanford Federal Facility Agreement and Consent Order, Milestone M-24-57 (2004). The need for the new well, well 299-W22-47, was identified during a data quality objectives process for establishing a RCRA/ Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA)/Atomic Energy Act (AEA) integrated 200 West and 200 East Area Groundwater Monitoring Network. This document provides a compilation of all available geologic data, spectral gamma ray logs, hydrogeologic data and well information obtained during drilling, well construction, well development, pump installation, aquifer testing, and sample collection/analysis activities. Appendix A contains the Well Summary Sheets, the Well Construction Summary Report, the geologist's Borehole Log, well development and pump installation records, and well survey results. Appendix B contains analytical results from groundwater samples collected during drilling. Appendix C contains complete spectral gamma ray logs and borehole deviation surveys.

  18. Waste Management Strategy for Dismantling Waste to Reduce Costs for Power Plant Decommissioning - 13543

    SciTech Connect (OSTI)

    Larsson, Arne; Lidar, Per [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)] [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden); Bergh, Niklas; Hedin, Gunnar [Westinghouse Electric Sweden AB, Fredholmsgatan 2, SE-721 63, Vaesteraas (Sweden)] [Westinghouse Electric Sweden AB, Fredholmsgatan 2, SE-721 63, Vaesteraas (Sweden)

    2013-07-01T23:59:59.000Z

    Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named 'ndcon' to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects. (authors)

  19. Thermal valorization of post-consumer film waste in a bubbling bed gasifier

    SciTech Connect (OSTI)

    Martínez-Lera, S., E-mail: susanamartinezlera@gmail.com; Torrico, J.; Pallarés, J.; Gil, A.

    2013-07-15T23:59:59.000Z

    Highlights: • Film waste from packaging is a common waste, a fraction of which is not recyclable. • Gasification can make use of the high energy value of the non-recyclable fraction. • This waste and two reference polymers were gasified in a bubbling bed reactor. • This experimental research proves technical feasibility of the process. • It also analyzes impact of composition and ER on the performance of the plant. - Abstract: The use of plastic bags and film packaging is very frequent in manifold sectors and film waste is usually present in different sources of municipal and industrial wastes. A significant part of it is not suitable for mechanical recycling but could be safely transformed into a valuable gas by means of thermal valorization. In this research, the gasification of film wastes has been experimentally investigated through experiments in a fluidized bed reactor of two reference polymers, polyethylene and polypropylene, and actual post-consumer film waste. After a complete experimental characterization of the three materials, several gasification experiments have been performed to analyze the influence of the fuel and of equivalence ratio on gas production and composition, on tar generation and on efficiency. The experiments prove that film waste and analogue polymer derived wastes can be successfully gasified in a fluidized bed reactor, yielding a gas with a higher heating value in a range from 3.6 to 5.6 MJ/m{sup 3} and cold gas efficiencies up to 60%.

  20. Emergency response packaging: A conceptual outline

    SciTech Connect (OSTI)

    Luna, Robert E.; McClure, J. D.; Bennett, P. C.; Wheeler, T. A.

    1991-01-01T23:59:59.000Z

    The main thrust of this paper has been to put forth the idea of developing a package for the recovery and retrieval of released radioactive material contents from Radioactive Materials (RAM) packaging involved in transport accidents. Prior to the development of such a package, some additional studies might be performed which would confirm the general type of candidate materials which might have to be recovered. This would require a detailed inventory of US packages that have released their contents due to transport accidents. The main issue is one of preparedness which would allow the US Department of Energy to respond to accidents for DOE shipments and to respond nationally for shipments outside the normal jurisdiction of US DOE shipments.

  1. Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2004-130

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-03-14T23:59:59.000Z

    The 1607-F1 Sanitary Sewer System (124-F-1), consisted of a septic tank, drain field, and associated pipelines that received sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office via the 100-F-26:8 pipelines. The septic tank required remedial action based on confirmatory sampling. 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.

  2. Remaining Sites Verification Package for the 600-111, P-11 Critical Mass Laboratory Crib, and UPR-600-16, Fire and Contamination Spread Waste Sites, Waste Site Reclassification Form 2004-065

    SciTech Connect (OSTI)

    J. M. Capron

    2008-10-28T23:59:59.000Z

    The 600-111, P-11 Critical Mass Laboratory Crib waste site, also referred to as the P-11 Facility, included the 120 Experimental Building, the 123 Control Building, and the P-11 Crib. The facility was constructed in 1949 and was used as a laboratory for plutonium criticality studies. In accordance with this evaluation, the confirmatory and verification sampling results support a reclassification of this site to Interim Closed Out. The results of confirmatory and 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.

  3. Concrete material characterization reinforced concrete tank structure Multi-Function Waste Tank Facility

    SciTech Connect (OSTI)

    Winkel, B.V.

    1995-03-03T23:59:59.000Z

    The purpose of this report is to document the Multi-Function Waste Tank Facility (MWTF) Project position on the concrete mechanical properties needed to perform design/analysis calculations for the MWTF secondary concrete structure. This report provides a position on MWTF concrete properties for the Title 1 and Title 2 calculations. The scope of the report is limited to mechanical properties and does not include the thermophysical properties of concrete needed to perform heat transfer calculations. In the 1970`s, a comprehensive series of tests were performed at Construction Technology Laboratories (CTL) on two different Hanford concrete mix designs. Statistical correlations of the CTL data were later generated by Pacific Northwest Laboratories (PNL). These test results and property correlations have been utilized in various design/analysis efforts of Hanford waste tanks. However, due to changes in the concrete design mix and the lower range of MWTF operating temperatures, plus uncertainties in the CTL data and PNL correlations, it was prudent to evaluate the CTL data base and PNL correlations, relative to the MWTF application, and develop a defendable position. The CTL test program for Hanford concrete involved two different mix designs: a 3 kip/in{sup 2} mix and a 4.5 kip/in{sup 2} mix. The proposed 28-day design strength for the MWTF tanks is 5 kip/in{sup 2}. In addition to this design strength difference, there are also differences between the CTL and MWTF mix design details. Also of interest, are the appropriate application of the MWTF concrete properties in performing calculations demonstrating ACI Code compliance. Mix design details and ACI Code issues are addressed in Sections 3.0 and 5.0, respectively. The CTL test program and PNL data correlations focused on a temperature range of 250 to 450 F. The temperature range of interest for the MWTF tank concrete application is 70 to 200 F.

  4. Cleanup Verification Package for the 118-F-6 Burial Ground

    SciTech Connect (OSTI)

    H. M. Sulloway

    2008-10-02T23:59:59.000Z

    This cleanup verification package documents completion of remedial action for the 118-F-6 Burial Ground located in the 100-FR-2 Operable Unit of the 100-F Area on the Hanford Site. The trenches received waste from the 100-F Experimental Animal Farm, including animal manure, animal carcasses, laboratory waste, plastic, cardboard, metal, and concrete debris as well as a railroad tank car.

  5. Nevada Test Site Waste Acceptance Criteria, December 2000

    SciTech Connect (OSTI)

    NONE

    2000-12-01T23:59:59.000Z

    This document establishes the US Department of Energy, Nevada Operations Office waste acceptance criteria. The waste acceptance criteria provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive waste and mixed waste for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Sites for storage or disposal.

  6. Development of Dodecaniobate Keggin Chain Materials as Alternative Sorbents for SR and Actinide Removal from High-Level Nuclear Waste Solutions

    SciTech Connect (OSTI)

    Nyman, May; Bonhomme, Francois

    2004-03-28T23:59:59.000Z

    The current baseline sorbent (monosodium titanate) for Sr and actinide removal from Savannah River Site's high level wastes has excellent adsorption capabilities for Sr but poor performance for the actinides. We are currently investigating the development of alternative materials that sorb radionuclides based on chemical affinity and/or size selectivity. The polyoxometalates, negatively-charged metal oxo clusters, have known metal binding properties and are of interest for radionuclide sequestration. We have developed a class of Keggin-ion based materials, where the Keggin ions are linked in 1- dimensional chains separated by hydrated, charge-balancing cations. These Nb-based materials are stable in the highly basic nuclear waste solutions and show good selectivity for Sr and Pu. Synthesis, characterization and structure of these materials in their native forms and Sr-exchanged forms will be presented.

  7. 1 INSTRODUCTION In the concept of geological radioactive waste disposal,

    E-Print Network [OSTI]

    Boyer, Edmond

    , thermal solicitation comes from the heat emitting from the radioactive waste packages. On one hand1 INSTRODUCTION In the concept of geological radioactive waste disposal, argillite is being elements from the waste package towards the environment. During the construction and exploitation phase

  8. Identification of lead chemical form in mine waste materials by X-ray absorption spectroscopy

    SciTech Connect (OSTI)

    Taga, Raijeli L.; Ng, Jack [University of Queensland, National Research Centre for Environmental Toxicology (EnTox), Brisbane, 4108 (Australia); Zheng Jiajia; Huynh, Trang; Noller, Barry [University of Queensland, Centre for Mined Land Rehabilitation, Brisbane, 4072 (Australia); Harris, Hugh H. [School of Chemistry and Physics, University of Adelaide, Adelaide, 5005 (Australia)

    2010-06-23T23:59:59.000Z

    X-ray absorption spectroscopy (XAS) provides a direct means for measuring lead chemical forms in complex samples. In this study, XAS was used to identify the presence of plumbojarosite (PbFe{sub 6}(SO{sub 4}){sub 4}(OH){sub 12}) by lead L{sub 3}-edge XANES spectra in mine waste from a small gold mining operation in Fiji. The presence of plumbojarosite in tailings was confirmed by XRD but XANES gave better resolution. The potential for human uptake of Pb from tailings was measured using a physiologically based extract test (PBET), an in-vitro bioaccessibility (BAc) method. The BAc of Pb was 55%. Particle size distribution of tailings indicated that 40% of PM{sub 10} particulates exist which could be a potential risk for respiratory effects via the inhalation route. Food items collected in the proximity of the mine site had lead concentrations which exceed food standard guidelines. Lead within the mining lease exceeded sediment guidelines. The results from this study are used to investigate exposure pathways via ingestion and inhalation for potential risk exposure pathways of Pb in that locality. The highest Pb concentration in soil and tailings was 25,839 mg/kg, exceeding the Australian National Environment Protection Measure (NEPM) soil health investigation levels.

  9. Remaining Sites Verification Package for the 600-111, P-11 Critical Mass Laboratory Crib, and UPR-600-16, Fire and Contamination Spread Waste Sites, Waste Site Reclassification Form 2008-045

    SciTech Connect (OSTI)

    J. M. Capron

    2008-10-28T23:59:59.000Z

    The UPR-600-16, Fire and Contamination Spread waste site is an unplanned release that occurred on December 4, 1951, when plutonium contamination was spread by a fire that ignited inside the 120 Experimental Building. The 120 Experimental Building was a laboratory building that was constructed in 1949 and used for plutonium criticality studies as part of the P-11 Project. In November 1951, a criticality occurred in the 120 Experimental Building that resulted in extensive plutonium contamination inside the building. The confirmatory evaluation supports 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 the extensive radiological survey of the surface soil and the confirmatory and 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. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

    SciTech Connect (OSTI)

    NONE

    1994-10-01T23:59:59.000Z

    The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results.

  11. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    SciTech Connect (OSTI)

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-12-31T23:59:59.000Z

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for {sup 238}Pu contaminated waste. Combustible low-level waste material contaminated with {sup 238}Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble {sup 2328}Pu in the spent salt. The valuable {sup 238}Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of {sup 238}Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered {sup 238}Pu is considered.

  12. Development of a tool dedicated to the evaluation of hydrogen term source for technological Wastes: assumptions, physical models, and validation

    SciTech Connect (OSTI)

    Lamouroux, C. [CEA Saclay, Nuclear Energy Division /DANS, Department of physico-chemistry, 91191 Gif sur yvette (France); Esnouf, S. [CEA Saclay, DSM/IRAMIS/SIS2M/Radiolysis Laboratory , 91191 Gif sur yvette (France); Cochin, F. [Areva NC,recycling BU, DIRP/RDP tour Areva, 92084 Paris La Defense (France)

    2013-07-01T23:59:59.000Z

    In radioactive waste packages hydrogen is generated, in one hand, from the radiolysis of wastes (mainly organic materials) and, in the other hand, from the radiolysis of water content in the cement matrix. In order to assess hydrogen generation 2 tools based on operational models have been developed. One is dedicated to the determination of the hydrogen source term issues from the radiolysis of the wastes: the STORAGE tool (Simulation Tool Of Emission Radiolysis Gas), the other deals with the hydrogen source term gas, produced by radiolysis of the cement matrices (the Damar tool). The approach used by the STORAGE tool for assessing the production rate of radiolysis gases is divided into five steps: 1) Specification of the data packages, in particular, inventories and radiological materials defined for a package medium; 2) Determination of radiochemical yields for the different constituents and the laws of behavior associated, this determination of radiochemical yields is made from the PRELOG database in which radiochemical yields in different irradiation conditions have been compiled; 3) Definition of hypothesis concerning the composition and the distribution of contamination inside the package to allow assessment of the power absorbed by the constituents; 4) Sum-up of all the contributions; And finally, 5) validation calculations by comparison with a reduced sampling of packages. Comparisons with measured values confirm the conservative character of the methodology and give confidence in the safety margins for safety analysis report.

  13. Cleanup Verification Package for the 618-8 Burial Ground

    SciTech Connect (OSTI)

    M. J. Appel

    2006-08-10T23:59:59.000Z

    This cleanup verification package documents completion of remedial action for the 618-8 Burial Ground, also referred to as the Solid Waste Burial Ground No. 8, 318-8, and the Early Solid Waste Burial Ground. During its period of operation, the 618-8 site is speculated to have been used to bury uranium-contaminated waste derived from fuel manufacturing, and construction debris from the remodeling of the 313 Building.

  14. Stabilizing plutonium materials at Hanford: systems engineering for PFP transition project effort on DNFSB 94-1

    SciTech Connect (OSTI)

    Huber, T.E., Westinghouse Hanford

    1996-07-02T23:59:59.000Z

    This report discusses the basic objectives of the stabilization and packaging activities at the Plutonium Finishing Plant that satisfy the Defense Nuclear Facility Safety Board Recommendation 94-1 by transforming the plutonium materials at hanford into forms or conditions which are suitable for safe storage to appropriate storage criteria; or discard that meets appropriate waste acceptance criteria.

  15. CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

    SciTech Connect (OSTI)

    Brinkman, K.; Fox, K.; Marra, J.

    2012-05-15T23:59:59.000Z

    The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets with CS/LN/TM combined waste stream with Mo and Zr removed. Waste streams that contain Mo must be produced in reducing environments to avoid Cs-Mo oxide phase formation. Waste streams without Mo have the ability to be melt processed in air. A path forward for further optimizing the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere, and controlled heat treatment schedules are anticipated to improve the targeted elemental partitioning.

  16. Management of Transuranic Contaminated Material

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1982-09-30T23:59:59.000Z

    To establish guidelines for the generation, treatment, packaging, storage, transportation, and disposal of transuranic (TRU) contaminated material.

  17. Status and Trend of Automotive Power Packaging

    SciTech Connect (OSTI)

    Liang, Zhenxian [ORNL

    2012-01-01T23:59:59.000Z

    Comprehensive requirements in aspects of cost, reliability, efficiency, form factor, weight, and volume for power electronics modules in modern electric drive vehicles have driven the development of automotive power packaging technology intensively. Innovation in materials, interconnections, and processing techniques is leading to enormous improvements in power modules. In this paper, the technical development of and trends in power module packaging are evaluated by examining technical details with examples of industrial products. The issues and development directions for future automotive power module packaging are also discussed.

  18. Gamma irradiation testing of montan wax barrier materials for in-situ waste containment

    SciTech Connect (OSTI)

    Soo, P.; Heiser, J.

    1996-02-01T23:59:59.000Z

    A scoping study was carried out to quantify the potential use of a montan wax as a barrier material for subsurface use. If it possesses resistance to chemical and structural change, it could be used in a barrier to minimize the migration of contaminants from their storage or disposal locations. Properties that were evaluated included hardness, melting point, molecular weight, and biodegradation as a function of gamma radiation dose. The main emphasis was to quantify the wax`s long-term ability to withstand radiation-induced mechanical, chemical, and microbial degradation.

  19. BALLISTICS TESTING OF THE 9977 SHIPPING PACKAGE FOR STORAGE APPLICATIONS

    SciTech Connect (OSTI)

    Loftin, B.; Abramczyk, G.; Koenig, R.

    2012-06-06T23:59:59.000Z

    Radioactive materials are stored in a variety of locations throughout the DOE complex. At the Savannah River Site (SRS), materials are stored within dedicated facilities. Each of those facilities has a documented safety analysis (DSA) that describes accidents that the facility and the materials within it may encounter. Facilities at the SRS are planning on utilizing the certified Model 9977 Shipping Package as a long term storage package and one of these facilities required ballistics testing. Specifically, in order to meet the facility DSA, the radioactive materials (RAM) must be contained within the storage package after impact by a .223 caliber round. In order to qualify the Model 9977 Shipping Package for storage in this location, the package had to be tested under these conditions. Over the past two years, the Model 9977 Shipping Package has been subjected to a series of ballistics tests. The purpose of the testing was to determine if the 9977 would be suitable for use as a storage package at a Savannah River Site facility. The facility requirements are that the package must not release any of its contents following the impact in its most vulnerable location by a .223 caliber round. A package, assembled to meet all of the design requirements for a certified 9977 shipping configuration and using simulated contents, was tested at the Savannah River Site in March of 2011. The testing was completed and the package was examined. The results of the testing and examination are presented in this paper.

  20. Thermoelectric generators incorporating phase-change materials for waste heat recovery from engine exhaust

    DOE Patents [OSTI]

    Meisner, Gregory P; Yang, Jihui

    2014-02-11T23:59:59.000Z

    Thermoelectric devices, intended for placement in the exhaust of a hydrocarbon fuelled combustion device and particularly suited for use in the exhaust gas stream of an internal combustion engine propelling a vehicle, are described. Exhaust gas passing through the device is in thermal communication with one side of a thermoelectric module while the other side of the thermoelectric module is in thermal communication with a lower temperature environment. The heat extracted from the exhaust gasses is converted to electrical energy by the thermoelectric module. The performance of the generator is enhanced by thermally coupling the hot and cold junctions of the thermoelectric modules to phase-change materials which transform at a temperature compatible with the preferred operating temperatures of the thermoelectric modules. In a second embodiment, a plurality of thermoelectric modules, each with a preferred operating temperature and each with a uniquely-matched phase-change material may be used to compensate for the progressive lowering of the exhaust gas temperature as it traverses the length of the exhaust pipe.

  1. Safety Analysis for Packaging Steel Banded Wooden Shipping Containers

    SciTech Connect (OSTI)

    FERRELL, P.C.

    2000-12-05T23:59:59.000Z

    This safety analysis report for packaging describes the steel banded wooden shipping containers, which are certified as Type AF packagings. The authorized payload for these containers is unirradiated, slightly enriched, uranium ingots, billets, extrusions, and scrap materials. The amount of uranium in the containers will not exceed the LSA-II material requirements as defined in 49 CFR 173.403.

  2. Waste characterization for the F/H Effluent Treatment Facility in support of waste certification

    SciTech Connect (OSTI)

    Brown, D.F.

    1994-10-17T23:59:59.000Z

    The Waste Acceptance Criteria (WAC) procedures define the rules concerning packages of solid Low Level Waste (LLW) that are sent to the E-area vaults (EAV). The WACs tabulate the quantities of 22 radionuclides that require manifesting in waste packages destined for each type of vault. These quantities are called the Package Administrative Criteria (PAC). If a waste package exceeds the PAC for any radionuclide in a given vault, then specific permission is needed to send to that vault. To avoid reporting insignificant quantities of the 22 listed radionuclides, the WAC defines the Minimum Reportable Quantity (MRQ) of each radionuclide as 1/1000th of the PAC. If a waste package contains less than the MRQ of a particular radionuclide, then the package`s manifest will list that radionuclide as zero. At least one radionuclide has to be reported, even if all are below the MRQ. The WAC requires that the waste no be ``hazardous`` as defined by SCDHEC/EPA regulations and also lists several miscellaneous physical/chemical requirements for the packages. This report evaluates the solid wastes generated within the F/H Effluent Treatment Facility (ETF) for potential impacts on waste certification.

  3. DESTRUCTIVE EXAMINATION OF SHIPPING PACKAGE 9975-03431

    SciTech Connect (OSTI)

    Daugherty, W.

    2012-05-30T23:59:59.000Z

    Destructive and non-destructive examinations have been performed on specified components of shipping package 9975-03431. For those attributes that were also measured during the field surveillance, no significant changes were observed. All observations and test results met identified criteria, or were collected for information and trending purposes. Except for modest corrosion of the lead shield (which is typical of these packages following several years service), no evidence of a degraded condition was found in this package. The Savannah River Site (SRS) stores packages containing plutonium (Pu) materials in the KArea Complex (KAC). The Pu materials are packaged per the DOE 3013 Standard and stored within Model 9975 shipping packages in KAC. The KAC facility DSA (Document Safety Analysis) credits the Model 9975 package to perform several safety functions, including criticality prevention, impact resistance, containment, and fire resistance to ensure the plutonium materials remain in a safe configuration during normal and accident conditions. The Model 9975 package is expected to perform its safety function for at least 12 years from initial packaging. The DSA recognizes the degradation potential for the materials of package construction over time in the KAC storage environment and requires an assessment of materials performance to validate the assumptions of the analysis and ultimately predict service life. As part of the comprehensive Model 9975 package surveillance program, destructive examination of package 9975-03431 was performed following field surveillance in accordance with Reference. Field surveillance of the Model 9975 package in KAC included nondestructive examination of the drum, fiberboard, lead shield and containment vessels. Results of the field surveillance are provided in Attachment 1.

  4. Mixed waste: Proceedings

    SciTech Connect (OSTI)

    Moghissi, A.A.; Blauvelt, R.K.; Benda, G.A.; Rothermich, N.E. [eds.] [Temple Univ., Philadelphia, PA (United States). Dept. of Environmental Safety and Health

    1993-12-31T23:59:59.000Z

    This volume contains the peer-reviewed and edited versions of papers submitted for presentation a the Second International Mixed Waste Symposium. Following the tradition of the First International Mixed Waste Symposium, these proceedings were prepared in advance of the meeting for distribution to participants. The symposium was organized by the Mixed Waste Committee of the American Society of Mechanical Engineers. The topics discussed at the symposium include: stabilization technologies, alternative treatment technologies, regulatory issues, vitrification technologies, characterization of wastes, thermal technologies, laboratory and analytical issues, waste storage and disposal, organic treatment technologies, waste minimization, packaging and transportation, treatment of mercury contaminated wastes and bioprocessing, and environmental restoration. Individual abstracts are catalogued separately for the data base.

  5. advanced packaging technology: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    David Smukowski, Sensors in Motion Embedded Actuation and Sensing Via Multi-Material 3D Printing.m. 3D Printing Multi-Functionality: Mec Materials for 2.5D and 3D Packaging Systems...

  6. advanced packaging technologies: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    David Smukowski, Sensors in Motion Embedded Actuation and Sensing Via Multi-Material 3D Printing.m. 3D Printing Multi-Functionality: Mec Materials for 2.5D and 3D Packaging Systems...

  7. advanced electronic packages: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    David Smukowski, Sensors in Motion Embedded Actuation and Sensing Via Multi-Material 3D Printing.m. 3D Printing Multi-Functionality: Mec Materials for 2.5D and 3D Packaging Systems...

  8. Determination of the Porosity Surfaces of the Disposal Room Containing Various Waste Inventories for WIPP PA.

    SciTech Connect (OSTI)

    Park, Byoung; Hansen, Francis D.

    2005-07-01T23:59:59.000Z

    This report develops a series of porosity surfaces for the Waste Isolation Pilot Plant. The concept of a porosity surface was developed for performance assessment and comprises calculation of room closure as salt creep processes are mitigated by gas generation and back stress created by the waste packages within the rooms. The physical and mechanical characteristics of the waste packaging that has already been disposed--such as the pipe overpack--and new waste packaging--such as the advanced mixed waste compaction--are appreciably different than the waste form upon which the original compliance was based and approved. This report provides structural analyses of room closure with various waste inventories. All of the underlying assumptions pertaining to the original compliance certification including the same finite element code are implemented; only the material parameters describing the more robust waste packages are changed from the certified baseline. As modeled, the more rigid waste tends to hold open the rooms and create relatively more void space in the underground than identical calculations run on the standard waste packages, which underpin the compliance certification. The several porosity surfaces quantified within this report provide possible ranges of pressure and porosity for performance assessment analyses.3 Intentionally blank4 AcknowledgementsThis research is funded by WIPP programs administered by the U.S. Department of Energy. The authors would like to acknowledge the valuable contributions to this work provided by others. Dr. Joshua S. Stein helped explain the hand off between these finite element porosity surfaces and implementation in the performance calculations. Dr. Leo L. Van Sambeek of RESPEC Inc. helped us understand the concepts of room closure under the circumstances created by a rigid waste inventory. Dr. T. William Thompson and Tom W. Pfeifle provided technical review and Mario J. Chavez provided a Quality Assurance review. The paper has been improved by these individuals.Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy under Contract DE-AC04-94Al850005 Intentionally Blank6

  9. EARLY TESTS OF DRUM TYPE PACKAGINGS - THE LEWALLEN REPORT

    SciTech Connect (OSTI)

    Smith, A.

    2010-07-29T23:59:59.000Z

    The need for robust packagings for radioactive materials (RAM) was recognized from the earliest days of the nuclear industry. The U.S. Department of Energy (DOE) Rocky Flats Plant developed a packaging for shipment of Pu in the early 1960's, which became the U.S. Department of Transportation (DOT) 6M specification package. The design concepts were employed in other early packagings. Extensive tests of these at Savannah River Laboratory (now Savannah River National Laboratory) were performed in 1969 and 1970. The results of these tests were reported in 'Drum and Board-Type Insulation Overpacks of Shipping Packages for Radioactive Materials', by E. E. Lewallen. The Lewallen Report was foundational to design of subsequent drum type RAM packaging. This paper summarizes this important early study of drum type packagings. The Lewallen Report demonstrated the ability packagings employing drum and insulation board overpacks and engineered containment vessels to meet the Type B package requirements. Because of the results of the Lewallen Report, package designers showed high concern for thermal protection of 'Celotex'. Subsequent packages addressed this by following strategies like those recommended by Lewallen and by internal metal shields and supplemental, encapsulated insulation disks, as in 9975. The guidance provide by the Lewallen Report was employed in design of a large number of drum size packagings over the following three decades. With the increased public concern over transportation of radioactive materials and recognition of the need for larger margins of safety, more sophisticated and complex packages have been developed and have replaced the simple packagings developed under the Lewallen Report paradigm.

  10. Geothermal reservoir simulation to enhance confidence in predictions for nuclear waste disposal

    E-Print Network [OSTI]

    Kneafsey, Timothy J.; Pruess, Karsten; O'Sullivan, Michael J.; Bodvarsson, Gudmundur S.

    2002-01-01T23:59:59.000Z

    for high-level nuclear waste. Journal of Contaminantfor a Potential High-Level Nuclear Waste Repository at YuccaHeat Flow Near High-Level Nuclear Waste Packages Emplaced in

  11. Recovering Americium and Curium from Mark-42 Target Materials- New Processing Approaches to Enhance Separations and Integrate Waste Stream Disposition - 12228

    SciTech Connect (OSTI)

    Patton, Brad D.; Benker, Dennis; Collins, Emory D.; Mattus, Catherine H.; Robinson, Sharon M.; Wham, Robert M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01T23:59:59.000Z

    Oak Ridge National Laboratory (ORNL) is investigating flowsheets to enhance processing efficiencies and to address waste streams associated with recovery of americium (Am) and curium (Cm) from Mark-42 (Mk-42) target materials stored at ORNL. The objective of this work was to identify the most effective flowsheet with which to process the 104 Mk-42 oxide capsules holding a total of 80 g of plutonium (Pu), 190 g of Cm, 480 g of Am, and 5 kg of lanthanide (Ln) oxides for the recovery and purification of the Am/Cm for future use as feedstock for heavy actinide production. Studies were also conducted to solidify the process flowsheet waste streams for disposal. ORNL is investigating flowsheets to enhance processing efficiencies and address waste streams associated with recovery of Am and Cm from Mk-42 target materials stored at ORNL. A series of small-scale runs are being performed to demonstrate an improved process to recover Am/Cm and to optimize the separations of Ln fission products from the Am/Cm constituents. The first of these runs has been completed using one of the Am/Cm/Ln oxide capsules stored at ORNL. The demonstration run showed promising results with a Ln DF of 40 for the Am/Cm product and an Am/Cm DF of 75 for the Ln product. In addition, the total losses of Am, Cm, and Ln to the waste solvents and raffinates were very low at <0.2%, 0.02%, and 0.04%, respectively. However, the Ln-actinide separation was less than desired. For future Reverse TALSPEAK demonstration runs, several parameters will be adjusted (flow rates, the ratio of scrub to strip stages, etc.) to improve the removal of Ln from the actinides. The next step will also include scale-up of the processing flowsheet to use more concentrated solutions (15 g/L Ln versus 5 g/L) and larger volumes and to recycle the HDEHP solvent. This should improve the overall processing efficiency and further reduce losses to waste streams. Studies have been performed with simulated wastes to develop solidification processes for disposal of the secondary waste streams generated by this flowsheet. Formulations were successfully developed for all the waste simulants. Additional tests with actual waste will be the next step in this effort. Future plans are to process all of the remaining 103 capsules in storage at ORNL. A nine-capsule test is now under way to provide additional information to scale-up the process to a target 20-capsule batch size for future processing runs. (authors)

  12. The ZOOM minimization package

    SciTech Connect (OSTI)

    Fischler, Mark S.; Sachs, D.; /Fermilab

    2004-11-01T23:59:59.000Z

    A new object-oriented Minimization package is available for distribution in the same manner as CLHEP. This package, designed for use in HEP applications, has all the capabilities of Minuit, but is a re-write from scratch, adhering to modern C++ design principles. A primary goal of this package is extensibility in several directions, so that its capabilities can be kept fresh with as little maintenance effort as possible. This package is distinguished by the priority that was assigned to C++ design issues, and the focus on producing an extensible system that will resist becoming obsolete.

  13. Pressure Build-Up During the Fire Test in Type B(U) Packages Containing Water - 13280

    SciTech Connect (OSTI)

    Feldkamp, Martin; Nehrig, Marko; Bletzer, Claus; Wille, Frank [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44, 12205 Berlin (Germany)] [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44, 12205 Berlin (Germany)

    2013-07-01T23:59:59.000Z

    The safety assessment of packages for the transport of radioactive materials with content containing liquids requires special consideration. The main focus is on water as supplementary liquid content in Type B(U) packages. A typical content of a Type B(U) package is ion exchange resin, waste of a nuclear power plant, which is not dried, normally only drained. Besides the saturated ion exchange resin, a small amount of free water can be included in these contents. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific issues. An overview of these issues is provided. The physical and chemical compatibility of the content itself and the content compatibility with the packages materials must be demonstrated for the assessment. Regarding the mechanical resistance the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could for example be caused by radiolysis of the liquid and must be taken into account for the storage period. If the package is stressed by the total inner pressure, this pressure leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The total pressure in any calculation is the sum of partial pressures of different gases which can be caused by different effects. The pressure build-up inside the package caused by the regulatory thermal test (30 min at 800 deg. C), as part of the cumulative test scenario under accident conditions of transport is discussed primarily. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from beginning of the thermal test until cooling-down. In this case, while calculating the temperature distribution, conduction and radiation as well as evaporation and condensation during the associated process of transport have to be considered. This paper discusses limiting amounts of water inside the cask which could lead to unacceptable pressure and takes into account saturated steam as well as overheated steam. However, the difficulties of assessing casks containing wet content will be discussed. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. (authors)

  14. Fate and transport processes controlling the migration of hazardous and radioactive materials from the Area 5 Radioactive Waste Management Site (RWMS)

    SciTech Connect (OSTI)

    Estrella, R.

    1994-10-01T23:59:59.000Z

    Desert vadose zones have been considered as suitable environments for the safe and long-term isolation of hazardous wastes. Low precipitation, high evapotranspiration and thick unsaturated alluvial deposits commonly found in deserts make them attractive as waste disposal sites. The fate and transport of any contaminant in the subsurface is ultimately determined by the operating retention and transformation processes in the system and the end result of the interactions among them. Retention (sorption) and transformation are the two major processes that affect the amount of a contaminant present and available for transport. Retention processes do not affect the total amount of a contaminant in the soil system, but rather decrease or eliminate the amount available for transport at a given point in time. Sorption reactions retard the contaminant migration. Permanent binding of solute by the sorbent is also possible. These processes and their interactions are controlled by the nature of the hazardous waste, the properties of the porous media and the geochemical and environmental conditions (temperature, moisture and vegetation). The present study summarizes the available data and investigates the fate and transport processes that govern the migration of contaminants from the Radioactive Waste Management Site (RWMS) in Area 5 of the Nevada Test Site (NTS). While the site is currently used only for low-level radioactive waste disposal, past practices have included burial of material now considered hazardous. Fundamentals of chemical and biological transformation processes are discussed subsequently, followed by a discussion of relevant results.

  15. Waste drum refurbishment

    SciTech Connect (OSTI)

    Whitmill, L.J.

    1996-10-18T23:59:59.000Z

    Low-carbon steel, radioactive waste containers (55-gallon drums) are experiencing degradation due to moisture and temperature fluctuations. With thousands of these containers currently in use; drum refurbishment becomes a significant issue for the taxpayer and stockholders. This drum refurbishment is a non-intrusive, portable process costing between 1/2 and 1/25 the cost of repackaging, depending on the severity of degradation. At the INEL alone, there are an estimated 9,000 drums earmarked for repackaging. Refurbishing drums rather than repackaging can save up to $45,000,000 at the INEL. Based on current but ever changing WIPP Waste Acceptance Criteria (WAC), this drum refurbishment process will restore drums to a WIPP acceptable condition plus; drums with up to 40% thinning o the wall can be refurbished to meet performance test requirements for DOT 7A Type A packaging. A refurbished drum provides a tough, corrosion resistant, waterproof container with longer storage life and an additional containment barrier. Drums are coated with a high-pressure spray copolymer material approximately .045 inches thick. Increase in internal drum temperature can be held to less than 15 F. Application can be performed hands-on or the equipment is readily adaptable and controllable for remote operations. The material dries to touch in seconds, is fully cured in 48 hours and has a service temperature of {minus}60 to 500 F. Drums can be coated with little or no surface preparation. This research was performed on drums however research results indicate the coating is very versatile and compatible with most any material and geometry. It could be used to provide abrasion resistance, corrosion protection and waterproofing to almost anything.

  16. Underground waste barrier structure

    DOE Patents [OSTI]

    Saha, Anuj J. (Hamburg, NY); Grant, David C. (Gibsonia, PA)

    1988-01-01T23:59:59.000Z

    Disclosed is an underground waste barrier structure that consists of waste material, a first container formed of activated carbonaceous material enclosing the waste material, a second container formed of zeolite enclosing the first container, and clay covering the second container. The underground waste barrier structure is constructed by forming a recessed area within the earth, lining the recessed area with a layer of clay, lining the clay with a layer of zeolite, lining the zeolite with a layer of activated carbonaceous material, placing the waste material within the lined recessed area, forming a ceiling over the waste material of a layer of activated carbonaceous material, a layer of zeolite, and a layer of clay, the layers in the ceiling cojoining with the respective layers forming the walls of the structure, and finally, covering the ceiling with earth.

  17. The basics in transportation of low-level radioactive waste

    SciTech Connect (OSTI)

    Allred, W.E.

    1998-06-01T23:59:59.000Z

    This bulletin gives a basic understanding about issues and safety standards that are built into the transportation system for radioactive material and waste in the US. An excellent safety record has been established for the transport of commercial low-level radioactive waste, or for that matter, all radioactive materials. This excellent safety record is primarily because of people adhering to strict regulations governing the transportation of radioactive materials. This bulletin discusses the regulatory framework as well as the regulations that set the standards for packaging, hazard communications (communicating the potential hazard to workers and the public), training, inspections, routing, and emergency response. The excellent safety record is discussed in the last section of the bulletin.

  18. Central Waste Complex (CWC) Waste Analysis Plan

    SciTech Connect (OSTI)

    ELLEFSON, M.D.

    1999-12-01T23:59:59.000Z

    The purpose of this waste analysis plan (WAP) is to document the waste acceptance process, sampling methodologies, analytical techniques, and overall processes that are undertaken for waste accepted for storage at the Central Waste Complex (CWC), which is located in the 200 West Area of the Hanford Facility, Richland, Washington. Because dangerous waste does not include the source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge.

  19. Protection of microelectronic devices during packaging

    DOE Patents [OSTI]

    Peterson, Kenneth A. (Albuquerque, NM); Conley, William R. (Tijeras, NM)

    2002-01-01T23:59:59.000Z

    The present invention relates to a method of protecting a microelectronic device during device packaging, including the steps of applying a water-insoluble, protective coating to a sensitive area on the device; performing at least one packaging step; and then substantially removing the protective coating, preferably by dry plasma etching. The sensitive area can include a released MEMS element. The microelectronic device can be disposed on a wafer. The protective coating can be a vacuum vapor-deposited parylene polymer, silicon nitride, metal (e.g. aluminum or tungsten), a vapor deposited organic material, cynoacrylate, a carbon film, a self-assembled monolayered material, perfluoropolyether, hexamethyldisilazane, or perfluorodecanoic carboxylic acid, silicon dioxide, silicate glass, or combinations thereof. The present invention also relates to a method of packaging a microelectronic device, including: providing a microelectronic device having a sensitive area; applying a water-insoluble, protective coating to the sensitive area; providing a package; attaching the device to the package; electrically interconnecting the device to the package; and substantially removing the protective coating from the sensitive area.

  20. Biohazardous Waste Disposal GuidelinesDescriptionStorage& LabelingTreatmentDisposal

    E-Print Network [OSTI]

    Wikswo, John

    Waste Sharps Waste Solid Lab Waste Liquid Waste Any of these devices if contaminated with biohazardousBiohazardous Waste Disposal GuidelinesDescriptionStorage& packaging LabelingTreatmentDisposal Mixed container. Container must be leakproof, ridgid, puncture resistant, clearly marked for biohazardous waste

  1. ADAPTING A CERTIFIED SHIPPING PACKAGE FOR STORAGE APPLICATIONS

    SciTech Connect (OSTI)

    Loftin, B.; Abramczyk, G.

    2012-06-05T23:59:59.000Z

    For years shipping packages have been used to store radioactive materials at many DOE sites. Recently, the K-Area Material Storage facility at the Savannah River Site became interested in and approved the Model 9977 Shipping Package for use as a storage package. In order to allow the 9977 to be stored in the facility, there were a number of evaluations and modifications that were required. There were additional suggested modifications to improve the performance of the package as a storage container that were discussed but not incorporated in the design that is currently in use. This paper will discuss the design being utilized for shipping and storage, suggested modifications that have improved the storage configuration but were not used, as well as modifications that have merit for future adaptations for both the 9977 and for other shipping packages to be used as storage packages.

  2. Transuranic (TRU) Waste Processing Center- Overview

    Broader source: Energy.gov [DOE]

    DOE established the TRU Waste Processing Center (TWPC) as a regional center for the management, treatment, packaging and shipment of DOE TRU waste legacy inventory. TWPC is also responsible for managing and treating Low Level and Mixed Low Level Waste generated at ORNL. TWPC is operated by Wastren Advantage, Inc. (WAI) under contract to the DOE's Oak Ridge Office.

  3. Safety evaluation for packaging (onsite) SERF cask

    SciTech Connect (OSTI)

    Edwards, W.S.

    1997-10-24T23:59:59.000Z

    This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  4. Quality of vacuum packaged lamb retail cuts

    E-Print Network [OSTI]

    Wanstedt, Kristen Gae

    1982-01-01T23:59:59.000Z

    ) provision of an ideal environment for the ag1ng of meat. Materials wh1ch prevent the rapid exchange of gases result in longer storage 11fe of meat than is obta1ned by packaging with more permeable materials (Kraft and Ayres, 1952). Jaye et al. (1962... storage for up to 35 days. Hanna et al. (1976) reported that these coryneform bacteria were species of ~Cba t i d a o hath i ~th rmos ha t m. B. ~th rmos hactu has been reported to be a major constituent of the bacterial flora of vacuum packaged...

  5. Cleanup Verification Package for the 118-F-1 Burial Ground

    SciTech Connect (OSTI)

    E. J. Farris and H. M. Sulloway

    2008-01-10T23:59:59.000Z

    This cleanup verification package documents completion of remedial action for the 118-F-1 Burial Ground on the Hanford Site. This burial ground is a combination of two locations formerly called Minor Construction Burial Ground No. 2 and Solid Waste Burial Ground No. 2. This waste site received radioactive equipment and other miscellaneous waste from 105-F Reactor operations, including dummy elements and irradiated process tubing; gun barrel tips, steel sleeves, and metal chips removed from the reactor; filter boxes containing reactor graphite chips; and miscellaneous construction solid waste.

  6. Nuclear Waste Management. Semiannual progress report, October 1984-March 1985

    SciTech Connect (OSTI)

    McElroy, J.L.; Powell, J.A. (comps.)

    1985-06-01T23:59:59.000Z

    Progress reports are presented for the following studies on radioactive waste management: defense waste technology; nuclear waste materials characterization center; and supporting studies. 19 figs., 29 tabs.

  7. Radiological Monitoring Results For Groundwater Samples Associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Pond: November 1, 2010-October 31, 2011

    SciTech Connect (OSTI)

    David Frederick

    2012-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed on samples from specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond (No.LA-000160-01). The radiological monitoring was performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  8. Radiological Monitoring Results For Groundwater Samples Associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Pond: May 1, 2010-October 31, 2010

    SciTech Connect (OSTI)

    David B. Frederick

    2011-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed on samples from specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond (#LA-000160-01). The radiological monitoring was performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  9. Radiological Monitoring Results for Groundwater Samples Associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Pond: November 1, 2012-October 31, 2013

    SciTech Connect (OSTI)

    Mike Lewis

    2014-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed on samples from specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond WRU-I-0160-01, Modification 1 (formerly LA-000160-01). The radiological monitoring was performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  10. Radiological Monitoring Results for Groundwater Samples Associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Pond: November 1, 2011-October 31, 2012

    SciTech Connect (OSTI)

    Mike lewis

    2013-02-01T23:59:59.000Z

    This report summarizes radiological monitoring performed on samples from specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit for the Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond WRU-I-0160-01, Modification 1 (formerly LA-000160-01). The radiological monitoring was performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  11. Harvesting nanoscale thermal radiation using pyroelectric materials

    E-Print Network [OSTI]

    Fang, Jin; Frederich, Hugo; Pilon, Laurent

    2010-01-01T23:59:59.000Z

    materials for harvesting waste heat”. International Journala way to convert waste heat directly into electricity. Ita novel way to harvest waste heat by combining pyroelectric

  12. DESTRUCTIVE EXAMINATION OF SHIPPING PACKAGE 9975-02028

    SciTech Connect (OSTI)

    Daugherty, W.; Stefek, T.

    2009-12-30T23:59:59.000Z

    Destructive and non-destructive examinations have been performed on specified components of shipping package 9975-02028. For those attributes that were also measured during the field surveillance, no significant changes were observed. Four conditions were identified that do not meet inspection criteria. These conditions are subject to additional investigation and disposition by the Surveillance Program Authority. The conditions include: (1) The lead shield was covered with a white corrosion layer; (2) The lead shield height exceeds drawing requirements; (3) Mold was observed on the lower fiberboard subassembly; and (4) Fiberboard thermal conductivity in the axial direction exceeded the specified range. The Surveillance Program Authority was notified of these conditions and will document the disposition by surveillance report. All other observations and test results met identified criteria, or were collected for information and trending purposes. The Savannah River Site (SRS) stores packages containing plutonium (Pu) materials in the K-Area Complex (KAC). The Pu materials are packaged per the DOE 3013 Standard and stored within Model 9975 shipping packages in KAC. The KAC facility DSA (Document Safety Analysis) credits the Model 9975 package to perform several safety functions, including criticality prevention, impact resistance, containment, and fire resistance to ensure the plutonium materials remain in a safe configuration during normal and accident conditions. The Model 9975 package is expected to perform its safety function for at least 12 years from initial packaging. The DSA recognizes the degradation potential for the materials of package construction over time in the KAC storage environment and requires an assessment of materials performance to validate the assumptions of the analysis and ultimately predict service life. As part of the comprehensive Model 9975 package surveillance program, destructive examination of package 9975-02028 was performed following field surveillance in accordance with Reference. Field surveillance of the Model 9975 package in KAC included nondestructive examination of the drum, fiberboard, lead shield and containment vessels. Results of the field surveillance are provided in Attachment 1.

  13. DESTRUCTIVE EXAMINATION OF SHIPPING PACKAGE 9975-02168

    SciTech Connect (OSTI)

    Daugherty, W.

    2010-11-18T23:59:59.000Z

    The Savannah River Site (SRS) stores packages containing plutonium (Pu) materials in the K-Area Complex (KAC). The Pu materials are packaged per the DOE 3013 Standard and stored within Model 9975 shipping packages in KAC. The KAC facility DSA (Document Safety Analysis) credits the Model 9975 package to perform several safety functions, including criticality prevention, impact resistance, containment, and fire resistance to ensure the plutonium materials remain in a safe configuration during normal and accident conditions. The Model 9975 package is expected to perform its safety function for at least 12 years from initial packaging. The DSA recognizes the degradation potential for the materials of package construction over time in the KAC storage environment and requires an assessment of materials performance to validate the assumptions of the analysis and ultimately predict service life. As part of the comprehensive Model 9975 package surveillance program, destructive examination of package 9975-02028 was performed following field surveillance in accordance with Reference. Field surveillance of the Model 9975 package in KAC included nondestructive examination of the drum, fiberboard, lead shield and containment vessels. Results of the field surveillance are provided in Attachment 1. Destructive and non-destructive examinations have been performed on specified components of shipping package 9975-02168. For those attributes that were also measured during the field surveillance, no significant changes were observed. Two conditions were identified that do not meet inspection criteria. These conditions are subject to additional investigation and disposition by the Surveillance Program Authority. The conditions include: (1) The lead shield was covered with a white corrosion layer, and (2) Fiberboard thermal conductivity in the axial direction exceeded the specified range. The Surveillance Program Authority was notified of these conditions and will document the findings by surveillance report. All other observations and test results met identified criteria, or were collected for information and trending purposes.

  14. Transportation and packaging resource guide

    SciTech Connect (OSTI)

    Arendt, J.W.; Gove, R.M.; Welch, M.J.

    1994-12-01T23:59:59.000Z

    The purpose of this resource guide is to provide a convenient reference document of information that may be useful to the U.S. Department of Energy (DOE) and DOE contractor personnel involved in packaging and transportation activities. An attempt has been made to present the terminology of DOE community usage as it currently exists. DOE`s mission is changing with emphasis on environmental cleanup. The terminology or nomenclature that has resulted from this expanded mission is included for the packaging and transportation user for reference purposes. Older terms still in use during the transition have been maintained. The Packaging and Transportation Resource Guide consists of four sections: Sect. 1, Introduction; Sect. 2, Abbreviations and Acronyms; Sect. 3, Definitions; and Sect. 4, References for packaging and transportation of hazardous materials and related activities, and Appendices A and B. Information has been collected from DOE Orders and DOE documents; U.S Department of Transportation (DOT), U.S. Environmental Protection Agency (EPA), and U.S. Nuclear Regulatory Commission (NRC) regulations; and International Atomic Energy Agency (IAEA) standards and other international documents. The definitions included in this guide may not always be a regulatory definition but are the more common DOE usage. In addition, the definitions vary among regulatory agencies. It is, therefore, suggested that if a definition is to be used in a regulatory or a legal compliance issue, the definition should be verified with the appropriate regulation. To assist in locating definitions in the regulations, a listing of all definition sections in the regulations are included in Appendix B. In many instances, the appropriate regulatory reference is indicated in the right-hand margin.

  15. Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario

    SciTech Connect (OSTI)

    Adkins, Harold E.; Cuta, Judith M.; Koeppel, Brian J.; Guzman, Anthony D.; Bajwa, Christopher S.

    2006-11-15T23:59:59.000Z

    On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB, the U.S. agency responsible for determining the cause of transportation accidents), to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation package designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the COBRA-SFS and ANSYS® computer codes to evaluate the thermal performance of different package designs. The staff concluded that larger transportation packages resembling the HOLTEC Model No. HI STAR 100 and TransNuclear Model No. TN-68 would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event with only minor damage to peripheral components. This is due to their sizable thermal inertia and design specifications in compliance with currently imposed regulatory requirements. The staff also concluded that some components of smaller transportation packages resembling the NAC Model No. LWT, despite placement within an ISO container, could degrade. USNRC staff evaluated the radiological consequences of the package responses to the Baltimore tunnel fire. Though components in some packages heated up beyond their service temperatures, the staff determined that there would be no significant dose as a result of the fire for any of these and similar packages.

  16. Understanding radioactive waste

    SciTech Connect (OSTI)

    Murray, R.L.

    1981-12-01T23:59:59.000Z

    This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

  17. Proceedings of the symposium on Scientific Basis for Nuclear Waste Management XXX

    SciTech Connect (OSTI)

    Dunn, Darrell [ed. Southwest Research Inst., San Antonio, Texas (United States); Poinssot, Christophe [ed. CEA-Saclay, 91191 Gif-sur-Yvette cedex (France); Begg, Bruce [ed. Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia)

    2007-07-01T23:59:59.000Z

    Management of nuclear waste remains an important international topic that includes reprocessing of commercial nuclear fuel, waste-form design and development, storage and disposal packaging, the process of repository site selection, system design, and performance assessment. Requirements to manage and dispose of materials from the production of nuclear weapons, and the renewed interest in nuclear power, in particular through the Generation IV Forum and the Advanced Fuel Cycle Initiative, can be expected to increase the need for scientific advances in waste management. A broad range of scientific and engineering disciplines is necessary to provide safe and effective solutions and address complex issues. This volume offers an interdisciplinary perspective on materials-related issues associated with nuclear waste management programs. Invited and contributed papers cover a wide range of topics including studies on: spent fuel; performance assessment and models; waste forms for low- and intermediate-level waste; ceramic and glass waste forms for plutonium and high-level waste; radionuclides; containers and engineered barriers; disposal environments and site characteristics; and partitioning and transmutation.

  18. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA

    SciTech Connect (OSTI)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION, NEVADA SITE OFFICE

    2005-07-01T23:59:59.000Z

    This document establishes the U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive and mixed waste for disposal. Mixed waste generated within the State of Nevada by NNSA/NSO activities is accepted for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Site for storage or disposal.

  19. Pyramiding tumuli waste disposal site and method of construction thereof

    DOE Patents [OSTI]

    Golden, Martin P. (Hamburg, NY)

    1989-01-01T23:59:59.000Z

    An improved waste disposal site for the above-ground disposal of low-level nuclear waste as disclosed herein. The disposal site is formed from at least three individual waste-containing tumuli, wherein each tumuli includes a central raised portion bordered by a sloping side portion. Two of the tumuli are constructed at ground level with adjoining side portions, and a third above-ground tumulus is constructed over the mutually adjoining side portions of the ground-level tumuli. Both the floor and the roof of each tumulus includes a layer of water-shedding material such as compacted clay, and the clay layer in the roofs of the two ground-level tumuli form the compacted clay layer of the floor of the third above-ground tumulus. Each tumulus further includes a shield wall, preferably formed from a solid array of low-level handleable nuclear wate packages. The provision of such a shield wall protects workers from potentially harmful radiation when higher-level, non-handleable packages of nuclear waste are stacked in the center of the tumulus.

  20. Evolution of Chemistry and Its Effects on the Corrosion of Engineered Barrier Materials

    SciTech Connect (OSTI)

    Dunn, Darrell [Mechanical and Materials Engineering, Southwest Research Institute, San Antonio, TX, 78238 (United States); Pan, Yi-Ming; He, Xihua; Yang, Lietai; Pabalan, Roberto [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX, 78238 (United States)

    2007-07-01T23:59:59.000Z

    The evolution of environmental conditions within the emplacement drifts of a potential high-level waste repository at Yucca Mountain, Nevada, may be influenced by several factors, including the temperature and relative humidity within the emplacement drifts and the composition of seepage water. The performance of the waste package and the drip shield may be affected by the evolution of the environmental conditions within the emplacement drifts. In this study, tests evaluated the evolution of environmental conditions on the waste package surfaces and in the surrounding host rock. The tests were designed to (i) simulate the conditions expected within the emplacement drifts; (ii) measure the changes in near-field chemistry; and (iii) determine environmental influence on the performance of the engineered barrier materials. Results of tests conducted in this study indicate the composition of salt deposits was consistent with the initial dilute water chemistry. Salts and possibly concentrated calcium chloride brines may be more aggressive than either neutral or alkaline brines. (authors)