National Library of Energy BETA

Sample records for waste package fill

  1. Depleted uranium oxides and silicates as spent nuclear fuel waste package fill materials

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-09-10

    A new repository waste package (WP) concept for spent nuclear fuel (SNF) is being investigated that uses depleted uranium (DU) to improve performance and reduce the uncertainties of geological disposal of SNF. The WP would be filled with SNF and then filled with depleted uranium (DU) ({approximately}0.2 wt % {sup 235}U) dioxide (UO{sub 2}) or DU silicate-glass beads. Fission products and actinides can not escape the SNF UO{sub 2} crystals until the UO{sub 2} dissolves or is transformed into other chemical species. After WP failure, the DU fill material slows dissolution by three mechanisms: (1) saturation of AT groundwater with DU and suppression of SNF dissolution, (2) maintenance of chemically reducing conditions in the WP that minimize SNF solubility by sacrificial oxidation of DU from the +4 valence state, and (3) evolution of DU to lower-density hydrated uranium silicates. The fill expansion seals the WP from water flow. The DU also isotopically exchanges with SNF uranium as the SNF degrades to reduce long-term nuclear-criticality concerns.

  2. Waste Package Lifting Calculation

    SciTech Connect (OSTI)

    H. Marr

    2000-05-11

    The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation.

  3. 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)

    1995-01-01

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

  4. Tritium waste package

    DOE Patents [OSTI]

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

    1995-11-07

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

  5. WASTE PACKAGE TRANSPORTER DESIGN

    SciTech Connect (OSTI)

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

    1998-09-23

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

  6. Waste disposal package

    DOE Patents [OSTI]

    Smith, M.J.

    1985-06-19

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

  7. Radioactive waste disposal package

    DOE Patents [OSTI]

    Lampe, Robert F. (Bethel Park, PA)

    1986-01-01

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

  8. Naval Waste Package Design Report

    SciTech Connect (OSTI)

    M.M. Lewis

    2004-03-15

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

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

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03

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

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

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1998-01-01

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

  11. The reduction of packaging waste

    SciTech Connect (OSTI)

    Raney, E.A.; Hogan, J.J.; McCollom, M.L.; Meyer, R.J.

    1994-04-01

    Nationwide, packaging waste comprises approximately one-third of the waste disposed in sanitary landfills. the US Department of Energy (DOE) generated close to 90,000 metric tons of sanitary waste. With roughly one-third of that being packaging waste, approximately 30,000 metric tons are generated per year. The purpose of the Reduction of Packaging Waste project was to investigate opportunities to reduce this packaging waste through source reduction and recycling. The project was divided into three areas: procurement, onsite packaging and distribution, and recycling. Waste minimization opportunities were identified and investigated within each area, several of which were chosen for further study and small-scale testing at the Hanford Site. Test results, were compiled into five ``how-to`` recipes for implementation at other sites. The subject of the recipes are as follows: (1) Vendor Participation Program; (2) Reusable Containers System; (3) Shrink-wrap System -- Plastic and Corrugated Cardboard Waste Reduction; (4) Cardboard Recycling ; and (5) Wood Recycling.

  12. Reference waste package environment report

    SciTech Connect (OSTI)

    Glassley, W.E.

    1986-10-01

    One of three candidate repository sites for high-level radioactive waste packages is located at Yucca Mountain, Nevada, in rhyolitic tuff 700 to 1400 ft above the static water table. Calculations indicate that the package environment will experience a maximum temperature of {similar_to}230{sup 0}C at 9 years after emplacement. For the next 300 years the rock within 1 m of the waste packages will remain dehydrated. Preliminary results suggest that the waste package radiation field will have very little effect on the mechanical properties of the rock. Radiolysis products will have a negligible effect on the rock even after rehydration. Unfractured specimens of repository rock show no change in hydrologic characteristics during repeated dehydration-rehydration cycles. Fractured samples with initially high permeabilities show a striking permeability decrease during dehydration-rehydration cycling, which may be due to fracture healing via deposition of silica. Rock-water interaction studies demonstrate low and benign levels of anions and most cations. The development of sorptive secondary phases such as zeolites and clays suggests that anticipated rock-water interaction may produce beneficial changes in the package environment.

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

    SciTech Connect (OSTI)

    BOEHNKE, W.M.

    2000-07-13

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

  14. YUCCA MOUNTAIN WASTE PACKAGE CLOSURE SYSTEM

    SciTech Connect (OSTI)

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

    2005-08-26

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

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

    SciTech Connect (OSTI)

    Romano, T.

    1997-09-25

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

  16. Waste Package Design Methodology Report

    SciTech Connect (OSTI)

    D.A. Brownson

    2001-09-28

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

  17. CH Packaging Operations for High Wattage Waste

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2006-01-06

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

  18. Conceptual waste packaging options for deep borehole disposal

    SciTech Connect (OSTI)

    Su, Jiann -Cherng; Hardin, Ernest L.

    2015-07-01

    This report presents four concepts for packaging of radioactive waste for disposal in deep boreholes. Two of these are reference-size packages (11 inch outer diameter) and two are smaller (5 inch) for disposal of Cs/Sr capsules. All four have an assumed length of approximately 18.5 feet, which allows the internal length of the waste volume to be 16.4 feet. However, package length and volume can be scaled by changing the length of the middle, tubular section. The materials proposed for use are low-alloy steels, commonly used in the oil-and-gas industry. Threaded connections between packages, and internal threads used to seal the waste cavity, are common oilfield types. Two types of fill ports are proposed: flask-type and internal-flush. All four package design concepts would withstand hydrostatic pressure of 9,600 psi, with factor safety 2.0. The combined loading condition includes axial tension and compression from the weight of a string or stack of packages in the disposal borehole, either during lower and emplacement of a string, or after stacking of multiple packages emplaced singly. Combined loading also includes bending that may occur during emplacement, particularly for a string of packages threaded together. Flask-type packages would be fabricated and heat-treated, if necessary, before loading waste. The fill port would be narrower than the waste cavity inner diameter, so the flask type is suitable for directly loading bulk granular waste, or loading slim waste canisters (e.g., containing Cs/Sr capsules) that fit through the port. The fill port would be sealed with a tapered, threaded plug, with a welded cover plate (welded after loading). Threaded connections between packages and between packages and a drill string, would be standard drill pipe threads. The internal flush packaging concepts would use semi-flush oilfield tubing, which is internally flush but has a slight external upset at the joints. This type of tubing can be obtained with premium, low-profile threaded connections at each end. The internal-flush design would be suitable for loading waste that arrives from the originating site in weld-sealed, cylindrical canisters. Internal, tapered plugs with sealing filet welds would seal the tubing at each end. The taper would be precisely machined onto both the tubing and the plug, producing a metal-metal sealing surface that is compressed as the package is subjected to hydrostatic pressure. The lower plug would be welded in place before loading, while the upper plug would be placed and welded after loading. Conceptual Waste Packaging Options for Deep Borehole Disposal July 30, 2015 iv Threaded connections between packages would allow emplacement singly or in strings screwed together at the disposal site. For emplacement on a drill string the drill pipe would be connected directly into the top package of a string (using an adapter sub to mate with premium semi-flush tubing threads). Alternatively, for wireline emplacement the same package designs could be emplaced singly using a sub with wireline latch, on the upper end. Threaded connections on the bottom of the lowermost package would allow attachment of a crush box, instrumentation, etc.

  19. CERAMIC WASTE FORM DATA PACKAGE

    SciTech Connect (OSTI)

    Amoroso, J.; Marra, J.

    2014-06-13

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

  20. Waste Package Component Design Methodology Report

    SciTech Connect (OSTI)

    D.C. Mecham

    2004-07-12

    This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.

  1. Performance assessment model of a single waste package (Conference...

    Office of Scientific and Technical Information (OSTI)

    WASTES; UNDERGROUND DISPOSAL; PACKAGING; DIFFUSION; GROUND WATER; RADIONUCLIDE MIGRATION; TECHNETIUM 99; SOLUBILITY 052002; 053004; WASTE DISPOSAL AND STORAGE; DESIGN BASIS ...

  2. Waste Package Materials Performance Peer Review

    Broader source: Energy.gov [DOE]

    A consensus peer review of the current technical basis and the planned experimental and modeling program for the prediction of the long-term performance of waste package materials being considered...

  3. Aqueous Corrosion Rates for Waste Package Materials

    SciTech Connect (OSTI)

    S. Arthur

    2004-10-08

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

  4. WASTE PACKAGE REMEDIATION SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    N.D. Sudan

    2000-06-22

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

  5. Nuclear waste package fabricated from concrete

    SciTech Connect (OSTI)

    Pfeiffer, P.A.; Kennedy, J.M.

    1987-03-01

    After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400/sup 0/C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs.

  6. Drift emplaced waste package thermal response

    SciTech Connect (OSTI)

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

    1993-01-01

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

  7. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    SciTech Connect (OSTI)

    S. LeStrange

    1999-11-15

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

  8. Industrial Waste Landfill IV upgrade package

    SciTech Connect (OSTI)

    Not Available

    1994-03-29

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

  9. Vitrification of waste with conitnuous filling and sequential melting

    DOE Patents [OSTI]

    Powell, James R.; Reich, Morris

    2001-09-04

    A method of filling a canister with vitrified waste starting with a waste, such as high-level radioactive waste, that is cooler than its melting point. Waste is added incrementally to a canister forming a column of waste capable of being separated into an upper zone and a lower zone. The minimum height of the column is defined such that the waste in the lower zone can be dried and melted while maintaining the waste in the upper zone below its melting point. The maximum height of the column is such that the upper zone remains porous enough to permit evolved gases from the lower zone to flow through the upper zone and out of the canister. Heat is applied to the waste in the lower zone to first dry then to raise and maintain its temperature to a target temperature above the melting point of the waste. Then the heat is applied to a new lower zone above the melted waste and the process of adding, drying and melting the waste continues upward in the canister until the entire canister is filled and the entire contents are melted and maintained at the target temperature for the desired period. Cooling of the melted waste takes place incrementally from the bottom of the canister to the top, or across the entire canister surface area, forming a vitrified product.

  10. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2005-04-04

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

  11. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2005-04-13

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

  12. THERMAL EVALUATION OF THE 2-MCO/2-DHLW WASTE PACKAGE

    SciTech Connect (OSTI)

    T. Schmitt

    2000-07-05

    The objective of this calculation was to determine the structural response of multi-canister overpacks (MCO) and the 2-MCO/2-Defense High-Level Waste (DHLW) Waste Package (WP) subjected to tip-over onto an unyielding surface (US). The scope of this calculation was limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design and was performed by the Waste Package Design Section in accordance with the DOE SNF Analysis Plan for FY 2000.

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

    SciTech Connect (OSTI)

    K.G. Mon

    2004-10-01

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

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

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

    2011-08-09

    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/supersede other directives.

  15. Safety evaluation for packaging for 1720-DR sodium-filled tank

    SciTech Connect (OSTI)

    Mercado, M.S.

    1996-03-09

    Preparations are under way to sell the sodium stored in the 1720-DR tank in the 1720-DR building. This will require that the tank, as well as the 1720-DR facility, be moved to the 300 Area, so that the sodium may be melted and transferred into a railroad tanker car. Because the sodium is a hazardous material and is being shipped in a nonspecification packaging, a safety evaluation for packaging (SEP) is required. This SEP approves the sodium-filled tank for a single shipment from the 105-DR area to the 300 Area.

  16. Thermal Evaluation of the Fort Saint Vrain Codisposal Waste Package

    SciTech Connect (OSTI)

    Adam Scheider; Horia Radulescu

    2001-07-19

    The objective of this calculation is to evaluate the thermal response of the Fort Saint Vrain (FSV) Codisposal Waste Package (WP) design under nominal Monitored Geologic Repository conditions. The objective of the calculation is to provide thermal parameter information to support the FSV waste package design. The information provided by the sketches (Attachment IV) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.124, ''Calculations'' (Ref. 17) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the SDHLW (Defense High Level Waste) / DOE (Department of Energy) Long WP.

  17. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    V. Delabrosse

    2003-02-27

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  18. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-29

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  19. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-05-06

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

  20. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-08-28

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

  1. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-03-21

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

  2. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2002-10-17

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

  3. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2002-12-18

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

  4. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-05-31

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

  5. 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.

    2011-08-31

    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

  6. Measurement of radionuclides in waste packages

    DOE Patents [OSTI]

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

    1986-01-01

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

  7. Measurement of radionuclides in waste packages

    DOE Patents [OSTI]

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

    1984-09-12

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

  8. Swing-Down of 21-PWR Waste Package

    SciTech Connect (OSTI)

    A.K. Scheider

    2001-05-04

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design.

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

    SciTech Connect (OSTI)

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

    1983-03-01

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

  10. Secondary Waste Form Down Selection Data Package ā€“ Ceramicrete

    SciTech Connect (OSTI)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31

    As part of high-level waste pretreatment and immobilized low activity waste processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed in the Integrated Disposal Facility. Currently, four waste forms are being considered for stabilization and solidification of the liquid secondary wastes. These waste forms are Cast Stone, Ceramicrete, DuraLith, and Fluidized Bed Steam Reformer. The preferred alternative will be down selected from these four waste forms. Pacific Northwest National Laboratory is developing data packages to support the down selection process. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilization and solidification of the liquid secondary wastes. The information included will be based on information available in the open literature and from data obtained from testing currently underway. This data package is for the Ceramicrete waste form. Ceramicrete is a relatively new engineering material developed at Argonne National Laboratory to treat radioactive and hazardous waste streams (e.g., Wagh 2004; Wagh et al. 1999a, 2003; Singh et al. 2000). This cement-like waste form can be used to treat solids, liquids, and sludges by chemical immobilization, microencapsulation, and/or macroencapsulation. The Ceramicrete technology is based on chemical reaction between phosphate anions and metal cations to form a strong, dense, durable, low porosity matrix that immobilizes hazardous and radioactive contaminants as insoluble phosphates and microencapsulates insoluble radioactive components and other constituents that do not form phosphates. Ceramicrete is a type of phosphate-bonded ceramic, which are also known as chemically bonded phosphate ceramics. The Ceramicrete 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.

  11. Long term nuclear criticality potential in waste packages

    SciTech Connect (OSTI)

    Thomas, D.A.; Doering, T.W.

    1994-12-31

    Title 10 CFR 60.131.(b).(7) requires that the radioactive waste disposed of in the Mined Geologic Disposal System (MGDS) remain subcritical during the period of isolation. The period of waste isolation, approximately 10,000 years, represents a time period greater than any previously examined for criticality control of spent fuel. Change in the criticality potential over long time periods for the Multi-Purpose Canister (MPC) waste package conceptual design has been examined and methods of criticality control over this time have been investigated.

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

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20

    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.

  13. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect (OSTI)

    J.M. Scaglione

    2004-12-17

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation.

  14. THERMAL ANALYSIS OF GEOLOGIC HIGH-LEVEL RADIOACTIVE WASTE PACKAGES

    SciTech Connect (OSTI)

    Hensel, S.; Lee, S.

    2010-04-20

    The engineering design of disposal of the high level waste (HLW) packages in a geologic repository requires a thermal analysis to provide the temperature history of the packages. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal gallery system and as input to assess the transient thermal characteristics of the vitrified HLW Package. The objective of the work was to evaluate the thermal performance of the supercontainer containing the vitrified HLW in a non-backfilled and unventilated underground disposal gallery. In order to achieve the objective, transient computational models for a geologic vitrified HLW package were developed by using a computational fluid dynamics method, and calculations for the HLW disposal gallery of the current Belgian geological repository reference design were performed. An initial two-dimensional model was used to conduct some parametric sensitivity studies to better understand the geologic system's thermal response. The effect of heat decay, number of co-disposed supercontainers, domain size, humidity, thermal conductivity and thermal emissivity were studied. Later, a more accurate three-dimensional model was developed by considering the conduction-convection cooling mechanism coupled with radiation, and the effect of the number of supercontainers (3, 4 and 8) was studied in more detail, as well as a bounding case with zero heat flux at both ends. The modeling methodology and results of the sensitivity studies will be presented.

  15. EXTERNAL CRITICALITY CALCULATION FOR DOE SNF CODISPOSAL WASTE PACKAGES

    SciTech Connect (OSTI)

    H. Radulescu

    2002-10-18

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

  16. Elemental characterization of LL-MA radioactive waste packages with the associated particle technique

    SciTech Connect (OSTI)

    Perot, B.; Carasco, C.; Toure, M.; El Kanawati, W.; Eleon, C.

    2011-07-01

    The French Alternative Energies and Atomic Energy Commission (CEA) and National Radioactive Waste Management Agency (ANDRA) are conducting an R and D program to improve the characterization of long-lived and medium activity (LL-MA) radioactive waste packages with analytical methods and with non-destructive nuclear measurements. This paper concerns fast neutron interrogation with the associated particle technique (APT), which brings 3D information about the waste material composition. The characterization of volume elements filled with iron, water, aluminium, and PVC in bituminized and fibre concrete LL-MA waste packages has been investigated with MCNP [1] and MODAR data analysis software [2]. APT provides usable information about major elements presents in the volumes of interest. However, neutron scattering on hydrogen nuclei spreads the tagged neutron beam out of the targeted volume towards surrounding materials, reducing spatial selectivity. Simulation shows that small less than 1 L targets can be characterised up to the half-radius of a 225 L bituminized drum, the matrix of which is very rich in hydrogen. Deeper characterization in concrete is possible but limited by counting statistics due to photon attenuation in this dense matrix and, unless large inspection volumes are considered, by the lack of spatial selectivity of the tagged neutron beam due to neutron scattering. (authors)

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

    SciTech Connect (OSTI)

    Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

    1990-12-01

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

  18. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-02-15

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

  19. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    SciTech Connect (OSTI)

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-10-03

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation.

  20. Thermal Evaluation for the Naval SNF Waste Package

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-04-25

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

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

    SciTech Connect (OSTI)

    Colleen Shelton-Davis

    2005-11-01

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

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

    SciTech Connect (OSTI)

    MANN, F.M.

    2000-03-02

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

  3. 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

    2013-07-01

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

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

    SciTech Connect (OSTI)

    Calmus, D.B.

    1994-08-25

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

  5. Method and apparatus for analyzing the fill characteristics of a packaging container

    DOE Patents [OSTI]

    Rodriguez, Julio G.

    1998-01-01

    A system for analyzing the fill characteristics of a container. A container having a filling material therein is positioned adjacent a sound generator. Sound waves from the generator are applied to the container, causing it to vibrate. A vibration detector is used to determine the amount of container vibration. A preferred vibration detector involves a laser vibrometer which applies a reference laser beam to the vibrating container. The reference beam is reflected off of the container to generate a reflected laser beam. The reflected beam experiences a Doppler frequency shift compared with the reference beam which is caused by container vibration. The Doppler shift of the reflected beam is then compared with standardized Doppler shift data from a control container. Repeated Doppler shift measurements may also be undertaken which are converted into a vibration profile that is compared with a standardized vibration profile from a control container.

  6. Method and apparatus for analyzing the fill characteristics of a packaging container

    DOE Patents [OSTI]

    Rodriguez, J.G.

    1998-10-13

    A system is described for analyzing the fill characteristics of a container. A container having a filling material therein is positioned adjacent a sound generator. Sound waves from the generator are applied to the container, causing it to vibrate. A vibration detector is used to determine the amount of container vibration. A preferred vibration detector involves a laser vibrometer which applies a reference laser beam to the vibrating container. The reference beam is reflected off of the container to generate a reflected laser beam. The reflected beam experiences a Doppler frequency shift compared with the reference beam which is caused by container vibration. The Doppler shift of the reflected beam is then compared with standardized Doppler shift data from a control container. Repeated Doppler shift measurements may also be undertaken which are converted into a vibration profile that is compared with a standardized vibration profile from a control container. 4 figs.

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

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2013-10-10

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

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

    SciTech Connect (OSTI)

    Westerman, R.E.

    1980-10-01

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

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

    SciTech Connect (OSTI)

    Ward, Andy L.

    2007-11-26

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

  10. 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.

    2002-02-26

    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.

  11. 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.

    2006-07-31

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

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

    SciTech Connect (OSTI)

    Norman E. Cole; Randy K. Elwood

    2003-10-01

    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.

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

    SciTech Connect (OSTI)

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

    2011-10-01

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

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

    SciTech Connect (OSTI)

    Horn, J. M., LLNL

    1998-01-01

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

  15. 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.

    1994-09-01

    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.

  16. 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

    1983-06-01

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

  17. TRANSPORT LOCOMOTIVE AND WASTE PACKAGE TRANSPORTER ITS STANDARDS IDENTIFICATION STUDY

    SciTech Connect (OSTI)

    K.D. Draper

    2005-03-31

    To date, the project has established important to safety (ITS) performance requirements for structures, systems and components (SSCs) based on identification and categorization of event sequences that may result in a radiological release. These performance requirements are defined within the ''Nuclear Safety Design Basis for License Application'' (NSDB) (BSC 2005). Further, SSCs credited with performing safe functions are classified as ITS. In turn, performance confirmation for these SSCs is sought through the use of consensus code and standards. The purpose of this study is to identify applicable codes and standards for the waste package (WP) transporter and transport locomotive ITS SSCs. Further, this study will form the basis for selection and the extent of applicability of each code and standard. This study is based on the design development completed for License Application only. Accordingly, identification of ITS SSCs beyond those defined within the NSDB are based on designs that may be subject to further development during detail design. Furthermore, several design alternatives may still be under consideration to satisfy certain safety functions, and that final selection will not be determined until further design development has occurred. Therefore, for completeness, throughout this study alternative designs currently under consideration will be discussed. Further, the results of this study will be subject to evaluation as part of a follow-on gap analysis study. Based on the results of this study the gap analysis will evaluate each code and standard to ensure each ITS performance requirement is fully satisfied. When a performance requirement is not fully satisfied a ''gap'' is highlighted. Thereafter, the study will identify supplemental requirements to augment the code or standard to meet performance requirements. Further, the gap analysis will identify non-standard areas of the design that will be subject to a Development Plan. Non-standard components and non-standard design configurations are defined as areas of the design that do not follow standard industry practices or codes and standards. Whereby, performance confirmation can not be readily sought through use of consensus standards.

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

    SciTech Connect (OSTI)

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

    2014-09-15

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

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

    SciTech Connect (OSTI)

    Kevin Skinner; Greg Housley; Colleen Shelton-Davis

    2011-11-01

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

  20. Thermal Response of the 44-BWR Waste Package to a Hypothetical Fire Accident

    SciTech Connect (OSTI)

    J.R. Smotrel; H. Marr; M.J. Anderson

    2001-04-05

    The purpose of this calculation is to determine the thermal response of the 44-boiling water reactor (BWR) waste package (WP) to the hypothetical regulatory fire accident. The objective is to calculate the temperature response of the waste package materials to the hypothetical short-term fire defined in 10 CFR 7 1, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the BWR thermal load are determined. The associated activity is the development of engineering evaluations to support the Licensing Application (LA) design activities.

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

    SciTech Connect (OSTI)

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

    2014-02-15

    Highlights: ā€¢ We study the recycling schemes of France, Germany, Portugal, Romania and the UK. ā€¢ The costs and benefits of recycling are compared for France, Portugal and Romania. ā€¢ The balance of costs and benefits depend on the perspective (strictly financial/economic). ā€¢ 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 ā€œrecycling systemā€ 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 ā€œpricesā€)

  2. Calculation of the Naval Long and Short Waste Package Three-Dimensional Thermal Interface Temperatures

    SciTech Connect (OSTI)

    H. Marr

    2006-10-25

    The purpose of this calculation is to evaluate the thermal performance of the Naval Long and Naval Short spent nuclear fuel (SNF) waste packages (WP) in the repository emplacement drift. The scope of this calculation is limited to the determination of the temperature profiles upon the surfaces of the Naval Long and Short SNF waste package for up to 10,000 years of emplacement. The temperatures on the top of the outside surface of the naval canister are the thermal interfaces for the Naval Nuclear Propulsion Program (NNPP). The results of this calculation are intended to support Licensing Application design activities.

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

    SciTech Connect (OSTI)

    Adkins, H.E.

    1996-10-29

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

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

    SciTech Connect (OSTI)

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

    2013-05-01

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

  5. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End ImpactsĀ 

    SciTech Connect (OSTI)

    V. DeLa Brosse

    2003-03-27

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  6. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End ImpactsĀ 

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-17

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  7. Stress Corrosion Cracking Model for High Level Radioactive-Waste Packages

    SciTech Connect (OSTI)

    P. Andresen; G. Gordon; S. Lu

    2004-10-05

    A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain repository. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022 (Alloy 22), the environment is represented by aqueous brine films present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the tensile stress is principally from weld induced residual stress. SCC has historically been separated into ''initiation'' and ''propagation'' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding); or that develop from corrosion processes such as pitting or dissolution of inclusions. To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment to determine the time to through-wall penetration for the waste package. This paper presents the development of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS-N06022 in environments relevant to high level radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository. In addition, a seismic damage related SCC crack opening area density model is briefly described.

  8. 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.

    1982-11-01

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

  9. Safety analysis report for packaging a DOT 7A specification container for tritiated liquid wastes

    SciTech Connect (OSTI)

    Alford, E.

    1980-08-01

    This Safety Analysis Report for Packaging (SARP) was prepared in accordance with ERDA (DOE) Appendix 5201 for DOE/ALO review and approval of packaging of tritiated liquid wastes to be shipped from Sandia National Laboratories, Livermore, (SNLL) California. This report presents information pertinent to the construction of tritiated liquid waste shipping containers. It contains design and development considerations, explains tests and evaluations required to prove the container can withstand normal transportation conditions, and demonstrates that the Sandia container-and-radioactive-material shipment package is in compliance with DOE and Department of Transportation (DOT) safety requirements. An internal review of this SARP has been performed in compliance with the ERDA (DOE) Manual, 5201 Appendix V.

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

    SciTech Connect (OSTI)

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

    2008-07-01

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

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

    SciTech Connect (OSTI)

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

    1996-03-01

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

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

    SciTech Connect (OSTI)

    Ladkany, S.G.

    1993-12-31

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

  13. 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.

    1988-08-01

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

  14. Abstraction of Models for Pitting and Crevice Corrosion of Drip Shield and Waste Package Outer Barrier

    SciTech Connect (OSTI)

    K. Mon

    2001-08-29

    This analyses and models report (AMR) was conducted in response to written work direction (CRWMS M and O 1999a). ICN 01 of this AMR was developed following guidelines provided in TWP-MGR-MD-000004 REV 01, ''Technical Work Plan for: Integrated Management of Technical Product Input Department'' (BSC 2001, Addendum B). The purpose and scope of this AMR is to review and analyze upstream process-level models (CRWMS M and O 2000a and CRWMS M and O 2000b) and information relevant to pitting and crevice corrosion degradation of waste package outer barrier (Alloy 22) and drip shield (Titanium Grade 7) materials, and to develop abstractions of the important processes in a form that is suitable for input to the WAPDEG analysis for long-term degradation of waste package outer barrier and drip shield in the repository. The abstraction is developed in a manner that ensures consistency with the process-level models and information and captures the essential behavior of the processes represented. Also considered in the model abstraction are the probably range of exposure conditions in emplacement drifts and local exposure conditions on drip shield and waste package surfaces. The approach, method, and assumptions that are employed in the model abstraction are documented and justified.

  15. Design and testing of Spec 7A containers for packaging radioactive wastes

    SciTech Connect (OSTI)

    Roberts, R.S.; Perkins, C.L.

    1982-11-19

    For a variety of reasons, the containers that have or currently are being used for packaging radioactive waste have drawbacks which has motivated LLNL to investigate, design and destructively test different Type A containers. The result of this work is manifested in the TX-4, which is comparatively lightweight, increases the net payload, and the simplicity of the design and ease in handling have proved to be timesaving. The TX-4 is readily available, relatively inexpensive and practical to use. It easily meets Type A packaging specifications with a gross payload of 7000 pounds. Although no tests were performed at a higher weight, we feel that the TX-4 could pass the tests at higher gross weights if the need arises. 20 figures.

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

    SciTech Connect (OSTI)

    NA

    2005-10-25

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

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

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-01-29

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

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

    SciTech Connect (OSTI)

    G. Gordon

    2004-10-13

    Stress corrosion cracking is one of the most common corrosion-related causes for premature breach of metal structural components. Stress corrosion cracking is the initiation and propagation of cracks in structural components due to three factors that must be present simultaneously: metallurgical susceptibility, critical environment, and static (or sustained) tensile stresses. This report was prepared according to ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of this report is to provide an evaluation of the potential for stress corrosion cracking of the engineered barrier system components (i.e., the drip shield, waste package outer barrier, and waste package stainless steel inner structural cylinder) under exposure conditions consistent with the repository during the regulatory period of 10,000 years after permanent closure. For the drip shield and waste package outer barrier, the critical environment is conservatively taken as any aqueous environment contacting the metal surfaces. Appendix B of this report describes the development of the SCC-relevant seismic crack density model (SCDM). The consequence of a stress corrosion cracking breach of the drip shield, the waste package outer barrier, or the stainless steel inner structural cylinder material is the initiation and propagation of tight, sometimes branching, cracks that might be induced by the combination of an aggressive environment and various tensile stresses that can develop in the drip shields or the waste packages. The Stainless Steel Type 316 inner structural cylinder of the waste package is excluded from the stress corrosion cracking evaluation because the Total System Performance Assessment for License Application (TSPA-LA) does not take credit for the inner cylinder. This document provides a detailed description of the process-level models that can be applied to assess the performance of Alloy 22 (used for the waste package outer barrier) and Titanium Grade 7 (used for the drip shield) that are subjected to the effects of stress corrosion cracking. The use of laser peening or other residual stress mitigation techniques is considered as a means of mitigating stress corrosion cracking in the waste package final closure lid weld.

  19. Assessment of dome-fill technology and potential fill materials for the Hanford single-shell tanks

    SciTech Connect (OSTI)

    Smyth, J.D.; Shade, J.W.; Somasundaram, S.

    1992-05-01

    This study is part of a task that will identify dome-fill materials to stabilize and prevent the collapse of the structures of 149 single- shell tanks (SSTs). The SSTs were built at the Hanford Site in Washington State and used between 1944 and 1980 to store radioactive and other hazardous wastes. In addition to identifying suitable fill materials, this task will develop the technology and methods required to fill the tanks with the selected material. To date, basalt is the only candidate fill material with any testing conducted for its suitability as a dome-fill material. Sufficient data do not exist to select or eliminate basalt as a candidate material. This report documents a review of past dome-fill work at the Hanford Site and of other pertinent literature to establish a baseline for the dome-fill technology. In addition, the report identifies existing dome-fill technology, preliminary performance criteria for dome-fill technology development, potential testing strategies, and potential fill materials. As a part of this study, potential fill materials are qualitatively evaluated and a list of preliminary candidate fill materials is identified. Future work will further screen these materials. The dome-fill task work will ultimately contribute to the development of a final waste form package and the safe isolation of wastes from the Hanford Site SSTs.

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

    SciTech Connect (OSTI)

    T. Wolery

    2005-02-22

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

  1. 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.

    1988-05-01

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

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

    SciTech Connect (OSTI)

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

    1986-03-01

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

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

    SciTech Connect (OSTI)

    PD Meyer; RJ Serne

    1999-12-21

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

  4. 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.

    2012-01-25

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

  5. A quantitative assessment of microbiological contributions to corrosion of candidate nuclear waste package materials

    SciTech Connect (OSTI)

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

    1999-07-01

    The US Department of Energy is contributing to the design of a potential nuclear waste repository at Yucca Mountain, Nevada. A system to predict the contribution of Yucca Mountain (YM) bacteria to overall corrosion rates of candidate waste package (WP) materials was designed and implemented. DC linear polarization resistance techniques were applied to candidate material coupons that had been inoculated with a mixture of YM-derived bacteria with potentially corrosive activities, or left sterile. Inoculated bacteria caused a 5- to 6-fold increase in corrosion rate of carbon steel C1020 (to approximately 7--8 {micro}m/yr), and an almost 100-fold increase in corrosion rate of Alloy 400 (to approximately 1 {micro}m/yr) was observed due to microbiological activities. Microbiologically Influenced Corrosion (MIC) rates on more resistant materials (CRMs: Alloy 625, Type 304 Stainless Steel, and Alloy C22) were on the order of hundredths of micrometers per year ({micro}m/yr). Bulk chemical and surfacial endpoint analyses of spent media and coupon surfaces showed preferential dissolution of nickel from Alloy 400 coupons and depletion of chromium from CRMs after incubation with YM bacteria. Scanning electron microscopy also showed greater damage to the Alloy 400 surface than that indicated by electrochemical detection methods.

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

    SciTech Connect (OSTI)

    H. Wang

    1997-01-23

    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.

  7. Integrating Volume Reduction and Packaging Alternatives to Achieve Cost Savings for Low Level Waste Disposal at the Rocky Flats Environmental Technology Site

    SciTech Connect (OSTI)

    Church, A.; Gordon, J.; Montrose, J. K.

    2002-02-26

    In order to reduce costs and achieve schedules for Closure of the Rocky Flats Environmental Technology Site (RFETS), the Waste Requirements Group has implemented a number of cost saving initiatives aimed at integrating waste volume reduction with the selection of compliant waste packaging methods for the disposal of RFETS low level radioactive waste (LLW). Waste Guidance Inventory and Shipping Forecasts indicate that over 200,000 m3 of low level waste will be shipped offsite between FY2002 and FY2006. Current projections indicate that the majority of this waste will be shipped offsite in an estimated 40,000 55-gallon drums, 10,000 metal and plywood boxes, and 5000 cargo containers. Currently, the projected cost for packaging, shipment, and disposal adds up to $80 million. With these waste volume and cost projections, the need for more efficient and cost effective packaging and transportation options were apparent in order to reduce costs and achieve future Site packaging a nd transportation needs. This paper presents some of the cost saving initiatives being implemented for waste packaging at the Rocky Flats Environmental Technology Site (the Site). There are many options for either volume reduction or alternative packaging. Each building and/or project may indicate different preferences and/or combinations of options.

  8. 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.

    1993-08-01

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

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

    SciTech Connect (OSTI)

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

    2000-01-18

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is known as the Hanford ILAW Performance Assessment (PA) Activity, hereafter called the ILAW PA project. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require predictions of contaminant migration from the facility. To make such predictions will require estimates of the fluxes of water moving through the sediments within the vadose zone around and beneath the disposal facility. These fluxes, loosely called recharge rates, are the primary mechanism for transporting contaminants to the groundwater. Pacific Northwest National Laboratory (PNNL) assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of recharge rates for current conditions and long-term scenarios involving the shallow-land disposal of ILAW. Specifically, recharge estimates are needed for a filly functional surface cover; the cover sideslope, and the immediately surrounding terrain. In addition, recharge estimates are needed for degraded cover conditions. The temporal scope of the analysis is 10,000 years, but could be longer if some contaminant peaks occur after 10,000 years. The elements of this report compose the Recharge Data Package, which provides estimates of recharge rates for the scenarios being considered in the 2001 PA. Table S.1 identifies the surface features and time periods evaluated. The most important feature, the surface cover, is expected to be the modified RCRA Subtitle C design. This design uses a 1-m-thick silt loam layer above sand and gravel filter layers to create a capillary break. A 0.15-m-thick asphalt layer underlies the filter layers to function as a backup barrier and to promote lateral drainage. Cover sideslopes are expected to be constructed with 1V:10H slopes using sandy gravel. The recharge estimates for each scenario were derived from lysimeter and tracer data collected by the ILAW PA and other projects and from modeling analyses.

  10. 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.

    2013-08-15

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

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

    SciTech Connect (OSTI)

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

    1986-09-30

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

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

    SciTech Connect (OSTI)

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

    1989-10-01

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

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

    SciTech Connect (OSTI)

    McCright, R.D.

    1985-09-30

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

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

    SciTech Connect (OSTI)

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

    1993-02-01

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

  15. 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.; Gordon, S.; Goldston, W.

    2013-07-08

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

  16. Tank Waste and Waste Processing | Department of Energy

    Office of Environmental Management (EM)

    Tank Waste and Waste Processing Tank Waste and Waste Processing Tank Waste and Waste Processing The Defense Waste Processing Facility set a record by producing 267 canisters filled ...

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

    SciTech Connect (OSTI)

    L.M. Montierth

    2000-09-15

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

  18. CH Packaging Operations Manual

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2005-06-13

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

  19. Remaining Sites Verification Package for 132-D-3, 1608-D Effluent Pumping Station, Waste Site Reclassification Form 2005-033

    SciTech Connect (OSTI)

    R. A. Carlson

    2006-05-09

    Decommissioning and demolition of the 132-D-3 site, 1608-D Effluent Pumping Station was performed in 1986. Decommissioning included removal of equipment, water, and sludge for disposal as radioactive waste. The at- and below-grade structure was demolished to at least 1 m below grade and the resulting rubble buried in situ. The area was backfilled to grade with at least 1 m of clean fill and contoured to the surrounding terrain. Residual concentrations support future land uses that can be represented by a rural-residential scenario and pose no threat to groundwater or the Columbia River based on RESRAD modeling.

  20. Data Package for Secondary Waste Form Down-Selectionā€”Cast Stone

    SciTech Connect (OSTI)

    Serne, R. Jeffrey; Westsik, Joseph H.

    2011-09-05

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

  1. Proactive Design of n-Type (In, Ce) Filled Skutterudites Enabling High-Temperature Waste Heat Recovery

    Broader source: Energy.gov [DOE]

    Thermoelectric and structural properties of n-type (In, Ce) filled skutterudites including power factors and ZT as a function of temperature are presented

  2. CH Packaging Operations Manual

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-06-26

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

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

    SciTech Connect (OSTI)

    Reidel, Steve P.; Chamness, Mickie A.

    2007-01-01

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

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

    SciTech Connect (OSTI)

    Colleen Shelton-Davis

    2003-09-01

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

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

    SciTech Connect (OSTI)

    Shelton-Davis, C.V.

    2003-09-26

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

  6. Influence of lithium-containing waste materials on the melting of packaging glass

    SciTech Connect (OSTI)

    Katkova, K.S.; Balandina, T.I.; Belyaeva, A.G.; Guloyan, Y.A.; Seregina, E.P.

    1986-07-01

    Lithium-containing waste materials from mica enrichment factories are studied. The possibilities of using the wastes for melting of green and semi-white sodium calcium silicate glasses are studied as well. Using physical methods, the authors study the influence of lithium-containing mica waste materials on the process of batch melting, silicate and glass formation, and clarification of molten glass. Tables show melting characteristics with various additions of lithium-containing mica waste, and the influence of added Li/sub 2/O on glass clarification. It is shown that the addition of lithium-containing mica waste materials to sodium calcium silicate glass has a positive effect on silicate and glass formation and on clarification.

  7. Enterprise Assessments Operational Awareness Record for the Review of the WTP Low-Activity Waste Facility Preliminary Documented Safety Analysis Change Package for the Effluent Management Facility (OAR # EA-WTP-LAW-2016-01-25)

    Broader source: Energy.gov [DOE]

    Operational Awareness Record for the Review of the Waste Treatment and Immobilization Plant Low-Activity Waste Facility Preliminary Documented Safety Analysis Change Package for the Effluent Management Facility

  8. Improving D&D Planning and Waste Management with Cutting and Packaging Simulation

    SciTech Connect (OSTI)

    Richard H. Meservey; Jean-Louis Bouchet

    2005-08-01

    The increased amount of decontamination and decommissioning (D&D) being performed throughout the world not only strains nuclear cleanup budgets, but places severe demands on the capacities of nuclear waste disposal sites. Although budgets and waste disposal sites have been able to accommodate the demand thus far, the increasing number of large facilities being decommissioned will cause major impacts to the waste disposal process. It is thus imperative that new and innovative technologies are applied within the D&D industry to reduce costs and waste disposal requirements for the decommissioning of our inventory of large and aging nuclear facilities. One of the most significant problems reactor ownerā€™s deal with is the accurate determination of the types and volumes of wastes that will be generated during decommissioning of their facilities. Waste disposal costs, restrictions, and transportation issues can account for as much as 30% of the total costs to decommission a facility and thus it is very important to have accurate waste volume estimates. The use of simulation technologies to estimate and reduce decommissioning waste volumes provides a new way to manage risks associated with this work. Simulation improves the process by allowing facility owners to obtain accurate estimates of the types and amounts of waste prior to starting the actual D&D work. This reduces risk by permitting earlier and better negotiations with the disposal sites, and more time to resolve transportation issues. While simulation is a tool to be used by the D&D contractors, its real value is in reducing risks and costs to the reactor owners.

  9. Remaining Sites Verification Package for the 100-F-50 Stormwater Runoff Culvert, Waste Site Reclassification Form 2007-001

    SciTech Connect (OSTI)

    J. M. Capron

    2008-04-15

    The 100-F-50 waste site, part of the 100-FR-2 Operable Unit, is a steel stormwater runoff culvert that runs between two railroad grades in the south-central portion of the 100-F Area. The culvert exiting the west side of the railroad grade is mostly encased in concrete and surrounded by a concrete stormwater collection depression partially filled with soil and vegetation. The drain pipe exiting the east side of the railroad grade embankment is partially filled with soil and rocks. The 100-F-50 stormwater diversion culvert confirmatory sampling results support a reclassification of this site to no action. The current site conditions achieve the remedial action objectives and corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  10. Degradation mode survey candidate titanium-base alloys for Yucca Mountain project waste package materials. Revision 1

    SciTech Connect (OSTI)

    Gdowski, G.E.

    1997-12-01

    The Yucca Mountain Site Characterization Project (YMP) is evaluating materials from which to fabricate high-level nuclear waste containers (hereafter called waste packages) for the potential repository at Yucca Mountain, Nevada. Because of their very good corrosion resistance in aqueous environments titanium alloys are considered for container materials. Consideration of titanium alloys is understandable since about one-third (in 1978) of all titanium produced is used in applications where corrosion resistance is of primary importance. Consequently, there is a considerable amount of data which demonstrates that titanium alloys, in general, but particularly the commercial purity and dilute {alpha} grades, are highly corrosion resistant. This report will discuss the corrosion characteristics of Ti Gr 2, 7, 12, and 16. The more highly alloyed titanium alloys which were developed by adding a small Pd content to higher strength Ti alloys in order to give them better corrosion resistance will not be considered in this report. These alloys are all two phase ({alpha} and {beta}) alloys. The palladium addition while making these alloys more corrosion resistant does not give them the corrosion resistance of the single phase {alpha} and near-{alpha} (Ti Gr 12) alloys.

  11. Literature review of intrinsic actinide colloids related to spent fuel waste package release rates

    SciTech Connect (OSTI)

    Zhao, P.; Steward, S.A.

    1997-01-01

    Existence of actinide colloids provides an important mechanism in the migration of radionuclides and will be important in performance of a geologic repository for high-level nuclear waste. Actinide colloids have been formed during long-term unsaturated dissolution of spent fuel by groundwater. This article summarizes a literature search of actinide colloids. This report emphasizes the formation of intrinsic actinide colloids, because they would have the opportunity to form soon after groundwater contact with the spent fuel and before actinide-bearing groundwater reaches the surrounding geologic formations.

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

    SciTech Connect (OSTI)

    Fayer, Michael J.; Keller, Jason M.

    2007-09-24

    Pacific Northwest National Laboratory (PNNL) assists CH2M HILL Hanford Group, Inc., in its preparation of the Resource Conservation and Recovery Act (RCRA) Facility Investigation report. One of the PNNL tasks is to use existing information to estimate recharge rates for past and current conditions as well as future scenarios involving cleanup and closure of tank farms. The existing information includes recharge-relevant data collected during activities associated with a host of projects, including those of RCRA, the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), the CH2M HILL Tank Farm Vadose Zone Project, and the PNNL Remediation and Closure Science Project. As new information is published, the report contents can be updated. The objective of this data package was to use published data to provide recharge estimates for the scenarios being considered in the RCRA Facility Investigation. Recharge rates were estimated for areas that remain natural and undisturbed, areas where the vegetation has been disturbed, areas where both the vegetation and the soil have been disturbed, and areas that are engineered (e.g., surface barrier). The recharge estimates supplement the estimates provided by PNNL researchers in 2006 for the Hanford Site using additional field measurements and model analysis using weather data through 2006.

  13. HOW THE ROCKY FLATS ENVIRONMENTAL TECHNOLOGY SITE DEVELOPED A NEW WASTE PACKAGE USING A POLYUREA COATING THAT IS SAFELY AND ECONOMICALLY ELIMINATING SIZE REDUCTION OF LARGE ITEMS

    SciTech Connect (OSTI)

    Dorr, Kent A.; Hogue, Richard S.; Kimokeo, Margaret K.

    2003-02-27

    One of the major challenges involved in closing the Rocky Flats Environmental Technology Site (RFETS) is the disposal of extremely large pieces of contaminated production equipment and building debris. Past practice has been to size reduce the equipment into pieces small enough to fit into approved, standard waste containers. Size reducing this equipment is extremely expensive, and exposes workers to high-risk tasks, including significant industrial, chemical, and radiological hazards. RFETS has developed a waste package using a Polyurea coating for shipping large contaminated objects. The cost and schedule savings have been significant.

  14. 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.

    2010-03-30

    This calculation package presents the results of an assessment of the performance of the 6 cell design of the Environmental Management Waste Management Facility (EMWMF). The calculations show that the new cell 6 design at the EMWMF meets the current WAC requirement. QA/QC steps were taken to verify the input/output data for the risk model and data transfer from modeling output files to tables and calculation.

  15. IN-PACKAGE CHEMISTRY ABSTRACTION

    SciTech Connect (OSTI)

    E. Thomas

    2005-07-14

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

  16. 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.; Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S.

    1990-06-01

    This report combines six work units performed in FY`85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs.

  17. A user's guide to the GoldSim/BLT-MS integrated software package:a low-level radioactive waste disposal performance assessment model.

    SciTech Connect (OSTI)

    Knowlton, Robert G.; Arnold, Bill Walter; Mattie, Patrick D.

    2007-03-01

    Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in the assessment of radioactive waste disposal and at the time of this publication is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. In countries with small radioactive waste programs, international technology transfer program efforts are often hampered by small budgets, schedule constraints, and a lack of experienced personnel. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available software codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission (NRC) and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, revitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a credible and solid computational platform for constructing probabilistic safety assessment models. This document is a reference users guide for the GoldSim/BLT-MS integrated modeling software package developed as part of a cooperative technology transfer project between Sandia National Laboratories and the Institute of Nuclear Energy Research (INER) in Taiwan for the preliminary assessment of several candidate low-level waste repository sites. Breach, Leach, and Transport-Multiple Species (BLT-MS) is a U.S. NRC sponsored code which simulates release and transport of contaminants from a subsurface low-level waste disposal facility. GoldSim is commercially available probabilistic software package that has radionuclide transport capabilities. The following report guides a user through the steps necessary to use the integrated model and presents a successful application of the paradigm of renewing legacy codes for contemporary application.

  18. In-Package Chemistry Abstraction

    SciTech Connect (OSTI)

    E. Thomas

    2004-11-09

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

  19. New_Package Example Package

    Energy Science and Technology Software Center (OSTI)

    2004-04-01

    A package created as a tool for developers wishing to Autotool (incorporate the Autotools configure and build processes into) a new or existing package. The package being AutotoolĀ’ed can be a Trilinos package, but New_Package is also more generally applicable. There is no useful functioning code in New_Package, The Autotools source files are extensively commented. The package has been used to assist developers in getting Trilinos packages converted to the Autotools configure and build processesmoreĀ Ā» in a short time.Ā«Ā less

  20. 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

    2005-12-21

    The 128-B-2 waste site was a burn pit historically used for the disposal of combustible and noncombustible wastes, including paint and solvents, office waste, concrete debris, and metallic debris. This site has been remediated by removing approximately 5,627 bank cubic meters of debris, ash, and contaminated soil to the Environmental Restoration Disposal Facility. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  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

    1996-06-12

    This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the Westinghouse Hanford Company, Waste Receiving and Processing Facility, Module 1 (WRAP-1) Drum. The WRAP-1 Drum was tested for DOE-HQ in August 1994, by Los Alamos National Laboratory, under docket number 94-35-7A. Additionally, comparison and evaluation of the approved, as-tested packaging configuration was performed by WHC in September 1995. The WRAP-1 Drum was evaluated against the performance of the DOT-17C, 208 1 (55-gal) steel drums tested and evaluated under dockets 89-13-7A/90-18-7A and 94-37-7A.

  2. Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area, Waste Site Reclassification Form 2006-003

    SciTech Connect (OSTI)

    R. A. Carlson

    2006-04-24

    The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  3. Review of SAR for Packaging Report

    Office of Environmental Management (EM)

    ... (and high-level-waste packages), their guidance ... with the maximum internal heat load and any residual ... caused by combustion or decomposition processes. ...

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

    SciTech Connect (OSTI)

    KELLY, D L

    2001-10-15

    This report documents the U.S. Department of Transportation Specification 7A Type A compliance test and evaluation results of the Standard Waste Box. Testing and evaluation activities documented herein are on behalf of the U.S. Department of Energy-Headquarters, Office of Safety, Health and Security (EM-5), Germantown, Maryland. Duratek Federal Services, Inc., Northwest Operations performed an evaluation of the changes as documented herein under Docket 01-53-7A.

  5. Data Package for Past and Current Groundwater Flow and Contamination beneath Single-Shell Tank Waste Management Areas

    SciTech Connect (OSTI)

    Horton, Duane G.

    2007-03-16

    This appendix summarizes historic and recent groundwater data collected from the uppermost aquifer beneath the 200 East and 200 West Areas. Although the area of interest is the Hanford Site Central Plateau, most of the information discussed in this appendix is at the scale of individual single-shell tank waste management areas. This is because the geologic, and thus the hydraulic, properties and the geochemical properties (i.e., groundwater composition) are different in different parts of the Central Plateau.

  6. Optimal segmentation and packaging process

    DOE Patents [OSTI]

    Kostelnik, Kevin M.; Meservey, Richard H.; Landon, Mark D.

    1999-01-01

    A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D&D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded.

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

    SciTech Connect (OSTI)

    KELLY, D.L.

    2001-08-15

    This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test and evaluation results of the Standard Waste Box (SWB). Testing and evaluation activities documented herein are on behalf of the U.S. Department of Energy-Headquarters (DOE-HQ), Office of Safety, Health and Security (EM-5), Germantown, Maryland. Dwatek Federal Services, Inc., Northwest Operations (DFSNW) performed an evaluation of the changes as documented herein under Docket 01-53-7A.

  8. Remaining Sites Verification Package for the 100-F-54 Animal Farm Pastures, Waste Site Reclassification Form 2008-015

    SciTech Connect (OSTI)

    J. M. Capron

    2008-04-17

    The 100-F-54 waste site, part of the 100-FR-2 Operable Unit, is the soil associated with the former pastures for holding domestic farm animals used in experimental toxicology studies. Evaluation of historical information resulted in identification of the experimental animal farm pastures as having potential residual soil contamination due to excrement from experimental animals. The 100-F-54 animal farm pastures confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  9. User Packages

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

    User Packages User Packages There is more than one way for users to manage installation of Python packages on their own. Users of the Anaconda distribution may create their own environments as described here. Users of the NERSC-built modules can use virtualenv, pip, or compile and install Python packages directly. Using "pip" on Edison and Cori There is an issue with the pip command on the Cray systems because of an SSL certificate verification problem. One symptom is that you create

  10. RH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2006-11-07

    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.

  11. RH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2008-01-12

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the "RH-TRU 72-B cask") and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a 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.

  12. Packaging and Transportation | Department of Energy

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

    Packaging and Transportation Packaging and Transportation Packaging and Transportation Radiological shipments are accomplished safely. Annually, about 400 million hazardous materials shipments occur in the United States by rail, air, sea, and land. Of these shipments, about three million are radiological shipments. Since Fiscal Year (FY) 2004, EM has completed over 150,000 shipments of radioactive material/waste. Please click here to see Office of Packaging and Transportation Fiscal Year 2012

  13. 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.

    2010-04-15

    The Transuranic (TRU) Disposition Project at Savannah River Site will require numerous transfers of radioactive materials within the site boundaries for sorting and repackaging. The three DOT Type A shipping packagings planned for this work have numerous bolts for securing the lids to the body of the packagings. In an effort to reduce operator time to open and close the packages during onsite transfers, thus reducing personnel exposure and costs, an evaluation was performed to analyze the effects of reducing the number of bolts required to secure the lid to the packaging body. The evaluation showed the reduction to one-third of the original number of bolts had no effect on the packagings capability to sustain vibratory loads, shipping loads, internal pressure loads, and the loads resulting from a 4-ft drop. However, the loads caused by the 4-ft drop are difficult to estimate and the study recommended each of the packages be dropped to show the actual effects on the package closure. Even with reduced bolting, the packagings were still required to meet the 49 CFR 178.350 performance criteria for Type A packaging. This paper discusses the effects and results of the drop testing of the three packagings.

  14. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2005-02-28

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

  15. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    None, None

    2009-06-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) Ā§71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are conducted. This document provides the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs.

  16. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    None, None

    2008-09-11

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the pplication." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) Ā§71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are conducted. This document provides the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs.

  17. Packaging and Transportation Support at LANL CTMA 2012

    SciTech Connect (OSTI)

    Salazar, Nick

    2012-06-08

    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.

  18. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2006-04-25

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package TransporterModel II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) §71.8. Any time a user suspects or has indications that the conditions ofapproval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are conducted.

  19. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    None, None

    2007-12-13

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) Ā§71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are conducted.

  20. Spack: the Supercomputing Package Manager

    Energy Science and Technology Software Center (OSTI)

    2013-11-09

    The HPC software ecosystem is growing larger and more complex, but software distribution mechanisms have not kept up with this trend. Tools, Libraries, and applications need to run on multiple platforms and build with multiple compliers. Increasingly, packages leverage common software components, and building any one component requires building all of its dependencies. In HPC environments, ABI-incompatible interfaces (likeMPI), binary-incompatible compilers, and cross-compiled environments converge to make the build process a combinatoric nightmare. This obstaclemoreĀ Ā» deters many users from adopting useful tools, and others waste countless hours building and rebuilding tools. Many package managers exist to solve these problems for typical desktop environments, but none suits the unique needs of supercomputing facilities or users. To address these problems, we have Spack, a package manager that eases the task of managing software for end-users, across multiple platforms, package versions, compilers, and ABI incompatibilities.Ā«Ā less

  1. Optimal segmentation and packaging process

    DOE Patents [OSTI]

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

    1999-08-10

    A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D and D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded. 3 figs.

  2. Norcal Waste Systems, Inc.

    SciTech Connect (OSTI)

    Not Available

    2002-12-01

    Fact sheet describes the LNG long-haul heavy-duty trucks at Norcal Waste Systems Inc.'s Sanitary Fill Company.

  3. WIPP Reaches Milestone Ā„ First Disposal Room Filled

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

    WIPP Reaches Milestone - First Disposal Room Filled CARLSBAD, N.M., September 4, 2001 - The U.S. Department of Energy's (DOE) Carlsbad Field Office today announced that Room 7 of Panel 1 at the Waste Isolation Pilot Plant (WIPP), the first underground room used for disposal operations, has been filled to capacity with transuranic waste. The milestone was reached at about 3:30 p.m. on August 24, as Waste Handling personnel emplaced a shipment of waste from the Idaho National Engineering and

  4. Proactive Design of n-Type (In, Ce) Filled Skutterudites Enabling...

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

    Design of n-Type (In, Ce) Filled Skutterudites Enabling High-Temperature Waste Heat Recovery Proactive Design of n-Type (In, Ce) Filled Skutterudites Enabling High-Temperature ...

  5. Pyrotechnic filled molding powder

    DOE Patents [OSTI]

    Hartzel, Lawrence W.; Kettling, George E.

    1978-01-01

    The disclosure relates to thermosetting molding compounds and more particularly to a pyrotechnic filled thermosetting compound comprising a blend of unfilled diallyl phthalate molding powder and a pyrotechnic mixture.

  6. 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

    2007-09-28

    This data package discusses the geochemistry of vadose zone sediments beneath the single-shell tank farms at the U.S. Department of Energyā€™s (DOEā€™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).

  7. Loose-fill insulations

    SciTech Connect (OSTI)

    1995-05-01

    Whether you are increasing the insulation levels in your current home or selecting insulation for a new home, choosing the right insulation material can be challenging. Fibrous loose-fill insulations such as cellulose, fiberglass, and rock wool are options you may wish to consider. This publication will introduce you to these materials--what they are, how they are applied, how they compare with each other, and other considerations regarding their use--so that you can decide whether loose fills are right for your home.

  8. Research and Development Program for transportation packagings at Sandia National Laboratories

    SciTech Connect (OSTI)

    Hohnstreiter, G.F.; Sorenson, K.B.

    1995-02-01

    This document contains information about the research and development programs dealing with waste transport at Sandia National Laboratories. This paper discusses topics such as: Why new packaging is needed; analytical methodologies and design codes;evaluation of packaging components; materials characterization; creative packaging concepts; packaging engineering and analysis; testing; and certification support.

  9. 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

    2008-03-03

    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.

  10. Legacy Waste | Department of Energy

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

    Services Ā» Legacy Waste Legacy Waste Legacy Waste The Environmental Management Los Alamos Field Office's (EM-LA) Solid Waste Stabilization and Disposition Project Team is dedicated to packaging, shipping and disposing legacy waste - low-level, mixed low-level, and transuranic (TRU) waste - from Los Alamos National Laboratory (LANL). The most challenging legacy waste at LANL is TRU, which is currently stored at Area G, located on a mesa 1.3 miles north of the residential community of White Rock

  11. Cost Estimation Package

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

    1997-03-28

    This chapter focuses on the components (or elements) of the cost estimation package and their documentation.

  12. Gas filled panel insulation

    DOE Patents [OSTI]

    Griffith, B.T.; Arasteh, D.K.; Selkowitz, S.E.

    1993-12-14

    A structural or flexible highly insulative panel which may be translucent, is formed from multi-layer polymeric material in the form of an envelope surrounding a baffle. The baffle is designed so as to minimize heat transfer across the panel, by using material which forms substantially closed spaces to suppress convection of the low conductivity gas fill. At least a portion of the baffle carries a low emissivity surface for suppression of infrared radiation. 18 figures.

  13. Gas filled panel insulation

    DOE Patents [OSTI]

    Griffith, Brent T.; Arasteh, Dariush K.; Selkowitz, Stephen E.

    1993-01-01

    A structural or flexible highly insulative panel which may be translucent, is formed from multi-layer polymeric material in the form of an envelope surrounding a baffle. The baffle is designed so as to minimize heat transfer across the panel, by using material which forms substantially closed spaces to suppress convection of the low conductivity gas fill. At least a portion of the baffle carries a low emissivity surface for suppression of infrared radiation.

  14. 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

    2008-11-07

    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.

  15. Remaining Sites Verification Package for the 100-F-26:14, 116-F-5 Influent Pipelines, Waste Site Reclassification Form 2007-029

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-02-29

    The 100-F-26:14 waste site includes underground pipelines associated with the 116-F-5 Ball Washer Crib and remnants of process pipelines on the west side of the 105-F Building. 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.

  16. Remaining Sites Verification Package for the 100-B-22:1 Pipelines and Associated Soils, Waste Site Reclassification Form 2005-042

    SciTech Connect (OSTI)

    L. M. Dittmer

    2006-09-12

    The 100-B-22:1 pipelines and associated soils were part of the 100-B Area water treatment facilities. The 100-B-22:1 waste site is limited to those pipelines that interconnected the 185-B Filter House, the 126-B-2 Clearwells, the 185-B Deaeration Plant, and the 190-B Process Pumphouse. None of the 100-B-22:1 pipelines carried environmentally significant contamination. In accordance with the historical information and field observations of this evaluation, the results support a reclassification of this site to No Action required to meet future rural-residential uses and be protective of groundwater and the Columbia River.

  17. Waste Isolation Pilot Plant Typifies Optimizing Resources to...

    Office of Environmental Management (EM)

    Typifies Optimizing Resources to Maximize Results Waste Isolation Pilot Plant Typifies ... HalfPACT transportation packages on a Waste Isolation Pilot Plant (WIPP) truck are ...

  18. Glass Ceramic Formulation Data Package

    SciTech Connect (OSTI)

    Crum, Jarrod V.; Rodriguez, Carmen P.; McCloy, John S.; Vienna, John D.; Chung, Chul-Woo

    2012-06-17

    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. 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 2005-004

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-03-14

    The 100-F-26:8 waste site consisted of the underground pipelines that conveyed sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office to the 1607-F1 septic tank. The site has been remediated and presently exists as an open excavation. 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.

  20. 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

    2008-05-30

    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.

  1. 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

    2008-04-29

    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.

  2. Dye filled security seal

    DOE Patents [OSTI]

    Wilson, Dennis C. W.

    1982-04-27

    A security seal for providing an indication of unauthorized access to a sealed object includes an elongate member to be entwined in the object such that access is denied unless the member is removed. The elongate member has a hollow, pressurizable chamber extending throughout its length that is filled with a permanent dye under greater than atmospheric pressure. Attempts to cut the member and weld it together are revealed when dye flows through a rupture in the chamber wall and stains the outside surface of the member.

  3. Permitting plan for the immobilized low-activity waste project

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-09-04

    This document addresses the environmental permitting requirements for the transportation and interim storage of the Immobilized Low-Activity Waste (ILAW) produced during Phase 1 of the Hanford Site privatization effort. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage and disposal of Tank Waste Remediation Systems (TWRS) immobilized low-activity tank waste (ILAW) and (2) interim storage of TWRS immobilized HLW (IHLW) and other canistered high-level waste forms. Low-activity waste (LAW), low-level waste (LLW), and high-level waste (HLW) are defined by the TWRS, Hanford Site, Richland, Washington, Final Environmental Impact Statement (EIS) DOE/EIS-0189, August 1996 (TWRS, Final EIS). By definition, HLW requires permanent isolation in a deep geologic repository. Also by definition, LAW is ``the waste that remains after separating from high-level waste as much of the radioactivity as is practicable that when solidified may be disposed of as LLW in a near-surface facility according to the NRC regulations.`` It is planned to store/dispose of (ILAW) inside four empty vaults of the five that were originally constructed for the Group Program. Additional disposal facilities will be constructed to accommodate immobilized LLW packages produced after the Grout Vaults are filled. The specifications for performance of the low-activity vitrified waste form have been established with strong consideration of risk to the public. The specifications for glass waste form performance are being closely coordinated with analysis of risk. RL has pursued discussions with the NRC for a determination of the classification of the Hanford Site`s low-activity tank waste fraction. There is no known RL action to change law with respect to onsite disposal of waste.

  4. Waste Isolation Pilot Plant simulated RH TRU waste experiments: Data and interpretation pilot

    SciTech Connect (OSTI)

    Molecke, M.A.; Argueello, G.J.; Beraun, R.

    1993-04-01

    The simulated, i.e., nonradioactive remote-handled transuranic waste (RH TRU) experiments being conducted underground in the Waste Isolation Pilot Plant (WIPP) were emplaced in mid-1986 and have been in heated test operation since 9/23/86. These experiments involve the in situ, waste package performance testing of eight full-size, reference RH TRU containers emplaced in horizontal, unlined test holes in the rock salt ribs (walls) of WIPP Room T. All of the test containers have internal electrical heaters; four of the test emplacements were filled with bentonite and silica sand backfill materials. We designed test conditions to be ``near-reference`` with respect to anticipated thermal outputs of RH TRU canisters and their geometrical spacing or layout in WIPP repository rooms, with RH TRU waste reference conditions current as of the start date of this test program. We also conducted some thermal overtest evaluations. This paper provides a: detailed test overview; comprehensive data update for the first 5 years of test operations; summary of experiment observations; initial data interpretations; and, several status; experimental objectives -- how these tests support WIPP TRU waste acceptance, performance assessment studies, underground operations, and the overall WIPP mission; and, in situ performance evaluations of RH TRU waste package materials plus design details and options. We provide instrument data and results for in situ waste container and borehole temperatures, pressures exerted on test containers through the backfill materials, and vertical and horizontal borehole-closure measurements and rates. The effects of heat on borehole closure, fracturing, and near-field materials (metals, backfills, rock salt, and intruding brine) interactions were closely monitored and are summarized, as are assorted test observations. Predictive 3-dimensional thermal and structural modeling studies of borehole and room closures and temperature fields were also performed.

  5. Hydrogen Filling Station

    SciTech Connect (OSTI)

    Boehm, Robert F; Sabacky, Bruce; Anderson II, Everett B; Haberman, David; Al-Hassin, Mowafak; He, Xiaoming; Morriseau, Brian

    2010-02-24

    Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. The Freedom CAR and Freedom FUEL initiatives emphasize the importance of hydrogen as a future transportation fuel. Presently, Las Vegas has one hydrogen fueling station powered by natural gas. However, the use of traditional sources of energy to produce hydrogen does not maximize the benefit. The hydrogen fueling station developed under this grant used electrolysis units and solar energy to produce hydrogen fuel. Water and electricity are furnished to the unit and the output is hydrogen and oxygen. Three vehicles were converted to utilize the hydrogen produced at the station. The vehicles were all equipped with different types of technologies. The vehicles were used in the day-to-day operation of the Las Vegas Valley Water District and monitoring was performed on efficiency, reliability and maintenance requirements. The research and demonstration utilized for the reconfiguration of these vehicles could lead to new technologies in vehicle development that could make hydrogen-fueled vehicles more cost effective, economical, efficient and more widely used. In order to advance the development of a hydrogen future in Southern Nevada, project partners recognized a need to bring various entities involved in hydrogen development and deployment together as a means of sharing knowledge and eliminating duplication of efforts. A road-mapping session was held in Las Vegas in June 2006. The Nevada State Energy Office, representatives from DOE, DOE contractors and LANL, NETL, NREL were present. Leadership from the National hydrogen Association Board of Directors also attended. As a result of this session, a roadmap for hydrogen development was created. This roadmap has the ability to become a tool for use by other road-mapping efforts in the hydrogen community. It could also become a standard template for other states or even countries to approach planning for a hydrogen future. Project partners also conducted a workshop on hydrogen safety and permitting. This provided an opportunity for the various permitting agencies and end users to gather to share experiences and knowledge. As a result of this workshop, the permitting process for the hydrogen filling station on the Las Vegas Valley Water Districtā€™s land was done more efficiently and those who would be responsible for the operation were better educated on the safety and reliability of hydrogen production and storage. The lessons learned in permitting the filling station and conducting this workshop provided a basis for future hydrogen projects in the region. Continuing efforts to increase the working pressure of electrolysis and efficiency have been pursued. Research was also performed on improving the cost, efficiency and durability of Proton Exchange Membrane (PEM) hydrogen technology. Research elements focused upon PEM membranes, electrodes/catalysts, membrane-electrode assemblies, seals, bipolar plates, utilization of renewable power, reliability issues, scale, and advanced conversion topics. Additionally, direct solar-to-hydrogen conversion research to demonstrate stable and efficient photoelectrochemistry (PEC) hydrogen production systems based on a number of optional concepts was performed. Candidate PEC concepts included technical obstacles such as inefficient photocatalysis, inadequate photocurrent due to non-optimal material band gap energies, rapid electron-hole recombination, reduced hole mobility and diminished operational lifetimes of surface materials exposed to electrolytes. Project Objective 1: Design, build, operate hydrogen filling station Project Objective 2: Perform research and development for utilizing solar technologies on the hydrogen filling station and convert two utility vehicles for use by the station operators Project Objective 3: Increase capacity of hydrogen filling station; add additional vehicle; conduct safety workshop; develop a roadmap for hydrogen development; accelerate the development of photovoltaic components Project Objective 4: Perform research on the Proton Exchange membrane

  6. PTS 13.2 Packaging and Preparation for Shipment 4/10/95

    Broader source: Energy.gov [DOE]

    The objective of this surveillance is to evaluate the effectiveness of the contractor's programs for packaging radioactive and hazardous wastes for shipment.Ā  The Facility Representative examines...

  7. 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.

    1985-10-01

    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.

  8. CAIRS Training Package

    Energy Savers [EERE]

    Direct Data Entry Training Package Version 7.0 June 2014 CAIRS Training Package (draft June 27, 2014) TABLE OF CONTENTS Introduction ..................................................................................................................................... 1 Business Rules for CAIRS Direct Data Entry ................................................................................ 2 CAIRS Case Input: Workspace vs. Production Space

  9. Didasko Tutorial Package

    Energy Science and Technology Software Center (OSTI)

    2005-06-09

    Didasko is the tutorial package of Trilinos. It contains several examples to explain the usage of the basic packages (in particular) Epetra, AztecOO, IFPACK, ML, Amesos, Teuchos, Triutils) and a PDF guide that details each example. No new algorithms are included in Didasko. This package is meant to be an introductory, self-contained reference for Trilinos users.

  10. Solid waste handling

    SciTech Connect (OSTI)

    Parazin, R.J.

    1995-05-31

    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.).

  11. Remaining Sites Verification Package for the 100-F-26:15 Miscellaneous Pipelines Associated with the 132-F-6, 1608-F Waste Water Pumping Station, Waste Site Reclassification Form 2007-031

    SciTech Connect (OSTI)

    L. M. Dittmer

    2008-03-18

    The 100-F-26:15 waste site consisted of the remnant portions of underground process effluent and floor drain pipelines that originated at the 105-F Reactor. 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.

  12. 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

    2008-03-14

    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.

  13. AGFA - Argonne Gas Filled Analyzer

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

    AGFA - Argonne Gas Filled Analyzer AGFA, the Argonne Gas-filled Fragment Analyzer is a state-of-the art gas-filled separator at ATLAS, which is being developed in collaboration among the Argonne Physics Division, Hebrew University, Jerusalem, University of Massachusetts, Lowell, University of Maryland, University of Edinburgh, Lawrence Berkeley National Laboratory and Oregon State University. This separator will be used for a wide range of studies, e.g. 1) in conjunction with Gammasphere for

  14. Lab sets new record for waste shipments

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

    uranium and above certain quantities. Liquid waste is not allowed. It's packaged in drums or boxes, which are then placed in a large, steel shipping container called a...

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

    Office of Environmental Management (EM)

    & Publications TEC Meeting Summaries - February 2008 Presentations Radioactive Waste Management Complex Wide Review Communication Is Key to Packaging and Transportation Safety...

  16. Packaged die heater

    DOE Patents [OSTI]

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

    2011-06-21

    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.

  17. Cleanup Verification Package for the 118-F-6 Burial Ground

    SciTech Connect (OSTI)

    H. M. Sulloway

    2008-10-02

    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.

  18. Stratimikos Wrapper Package

    Energy Science and Technology Software Center (OSTI)

    2006-08-22

    Stratimikos is a small package of C++ wrappers for linear solver and preconditioning functionality exposed through Thyra interfaces. This package makes is possible to aggregate all of the general linear solver capability from the packages Amesos, AztecOO, Belos, lfpack, ML and others into a simple to use, parameter-list driven, interface to linear solvers. This initial version of Stratimikos contains just one utility class for building linear solvrs and preconditioners out of Epetra-based linear operators.

  19. HVAC Packages for SMSCB

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

    HVAC Packages for SMSCB* 2015 Building Technologies Office Peer Review * Small and Medium Sized Commercial Buildings Russell D. Taylor, TaylorRD@utrc.utc.com CBEI - United ...

  20. DOT specification packages evaluation

    SciTech Connect (OSTI)

    Ratledge, J.E.; Rawl, R.R. )

    1991-01-01

    During the late 1960s and early 1970s, the Department of Transportation (DOT) specification package system was implemented to serve as a useful and equivalent alternative to the Nuclear Regulatory Commission (NRC) and the Bureau of Explosives approval systems for Type B and fissile radioactive material package designs. When a package design was used by a large number of organizations, the package design was added to the DOT regulations as a specification package authorized for use by any shipper. In the mid-1970s, the NRC revised its package design certification system to the one in use today. This paper reports that, while the NRC and DOT transportation regulations have evolved over the years, the DOT specification package designs have remained largely unchanged. Questions have been raised as to whether these designs meet the current and proposed regulations. In order to enable the NRC and DOT to develop a regulatory analysis that will support appropriate action regarding the specification packages, a study is being performed to compile all available design, testing, and analysis information on these packages.

  1. Safety analysis report for packaging (onsite) multicanister overpack cask

    SciTech Connect (OSTI)

    Edwards, W.S.

    1997-07-14

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  2. COMSOL MULTIPHYSICS MODEL FOR DWPF CANISTER FILLING

    SciTech Connect (OSTI)

    Kesterson, M.

    2011-03-31

    The purpose of this work was to develop a model that can be used to predict temperatures of the glass in the Defense Waste Processing Facility (DWPF) canisters during filling and cooldown. Past attempts to model these processes resulted in large (>200K) differences in predicted temperatures compared to experimentally measured temperatures. This work was therefore intended to also generate a model capable of reproducing the experimentally measured trends of the glass/canister temperature during filling and subsequent cooldown of DWPF canisters. To accomplish this, a simplified model was created using the finite element modeling software COMSOL Multiphysics which accepts user defined constants or expressions to describe material properties. The model results were compared to existing experimental data for validation. A COMSOL Multiphysics model was developed to predict temperatures of the glass within DWPF canisters during filling and cooldown. The model simulations and experimental data were in good agreement. The largest temperature deviations were {approx}40 C for the 87inch thermocouple location at 3000 minutes and during the initial cooldown at the 51 inch location occurring at approximately 600 minutes. Additionally, the model described in this report predicts the general trends in temperatures during filling and cooling observed experimentally. However, the model was developed using parameters designed to fit a single set of experimental data. Therefore, Q-loss is not currently a function of pour rate and pour temperature. Future work utilizing the existing model should include modifying the Q-loss term to be variable based on flow rate and pour temperature. Further enhancements could include eliminating the Q-loss term for a user defined convection where Navier-Stokes does not need to be solved in order to have convection heat transfer.

  3. FILL IN OF CONTRACT CLAUSES

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

    FILL IN OF CONTRACT CLAUSES Certain solicitation clauses have been uploaded with the solicitation, as a separate Word document (Clauses.doc), which are to be completed by the Offeror, saved and submit as File 3. All areas marked as TBD in these clauses are to be filled in by the Offeror, this information shall then be utilized to complete these specific areas prior to contract award. B.1 SERVICES BEING ACQUIRED - INFORMATION TECHNOLOGY OPERATIONS AND MAINTENANCE SUPPORT SERVICES The Contractor

  4. Seawater Chemistry Package

    Energy Science and Technology Software Center (OSTI)

    2005-11-23

    SeaChem Seawater Chemistry package provides routines to calculate pH, carbonate chemistry, density, and other quantities for seawater, based on the latest community standards. The chemistry is adapted from fortran routines provided by the OCMIP3/NOCES project, details of which are available at http://www.ipsl.jussieu.fr/OCMIP/. The SeaChem package can generate Fortran subroutines as well as Python wrappers for those routines. Thus the same code can be used by Python or Fortran analysis packages and Fortran ocean models alike.

  5. Nuclear Material Packaging

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

    2008-03-07

    The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. Supersedes DOE M 441.1-1, dated 3-7-08.

  6. Nuclear Material Packaging Manual

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

    2008-03-07

    The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. Does not cancel/supersede other directives. Certified 11-18-10.

  7. Packaging and Transportation Safety

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

    2010-05-14

    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. Supersedes DOE O 460.1B.

  8. Packaging and Transportation Safety

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

    1995-09-27

    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

  9. Packaging and Transportation Safety

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

    1995-09-27

    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

  10. Packaging and Transportation Safety

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

    1996-10-02

    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.

  11. Waste acceptance criteria for the Waste Isolation Pilot Plant. Revision 4

    SciTech Connect (OSTI)

    Not Available

    1991-12-01

    This Revision 4 of the Waste Acceptance Criteria (WAC), WIPP-DOE-069, identifies and consolidates existing criteria and requirements which regulate the safe handling and preparation of Transuranic (TRU) waste packages for transportation to and emplacement in the Waste Isolation Pilot Plant (WIPP). This consolidation does not invalidate any existing certification of TRU waste to the WIPP Operations and Safety Criteria (Revision 3 of WIPP-DOE--069) and/or Transportation: Waste Package Requirements (TRUPACT-II Safety Analysis Report for Packaging [SARP]). Those documents being consolidated, including Revision 3 of the WAC, currently support the Test Phase.

  12. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-04-30

    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 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 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 10 CFR 71.11. 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 provides the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs.

  13. CH Packaging Program Guidance

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2002-03-04

    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 TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event there is a conflict between this document and the SARP or C of C, the SARP and/or C of C shall govern. 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 SAR P charges the WIPP Management and Operation (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 10 CFR 71.11. 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 TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize these operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs.

  14. Safety Evaluation for Packaging (onsite) T Plant Canyon Items

    SciTech Connect (OSTI)

    OBRIEN, J.H.

    2000-07-14

    This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments.

  15. FFTF railroad tank car safety evaluation for packaging

    SciTech Connect (OSTI)

    Romano, T.

    1996-10-25

    This Safety Evaluation for Packaging (SEP) provides evaluations necessary to approve transfer of the 8,000 gallon Liquid Waste Tank Car (LWTC) from the Fast Flux Test Facility (FFTF) to the 200 Areas. This SEP will demonstrate that the transfer cif the LWTC will provide an equivalent degree of safety as would be provided by packages meeting U.S. Department of Transportation (DOT) requirements. This fulfills onsite transportation requirements implemented in the Hazardous Material Packaging and Shipping, WHC-CM-2-14.

  16. Detecting small holes in packages

    DOE Patents [OSTI]

    Kronberg, J.W.; Cadieux, J.R.

    1996-03-19

    A package containing a tracer gas, and a method for determining the presence of a hole in the package by sensing the presence of the gas outside the package are disclosed. The preferred tracer gas, especially for food packaging, is sulfur hexafluoride. A quantity of the gas is added to the package and the package is closed. The concentration of the gas in the atmosphere outside the package is measured and compared to a predetermined value of the concentration of the gas in the absence of the package. A measured concentration greater than the predetermined value indicates the presence of a hole in the package. Measuring may be done in a chamber having a lower pressure than that in the package. 3 figs.

  17. Detecting small holes in packages

    DOE Patents [OSTI]

    Kronberg, James W.; Cadieux, James R.

    1996-01-01

    A package containing a tracer gas, and a method for determining the presence of a hole in the package by sensing the presence of the gas outside the package. The preferred tracer gas, especially for food packaging, is sulfur hexafluoride. A quantity of the gas is added to the package and the package is closed. The concentration of the gas in the atmosphere outside the package is measured and compared to a predetermined value of the concentration of the gas in the absence of the package. A measured concentration greater than the predetermined value indicates the presence of a hole in the package. Measuring may be done in a chamber having a lower pressure than that in the package.

  18. Waste characterization for the F/H Effluent Treatment Facility in support of waste certification

    SciTech Connect (OSTI)

    Brown, D.F.

    1994-10-17

    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.

  19. TSF Interface Package

    Energy Science and Technology Software Center (OSTI)

    2004-03-01

    A collection of packages of classes for interfacing to sparse and dense matrices, vectors and graphs, and to linear operators. TSF (via TSFCore, TSFCoreUtils and TSFExtended) provides the application programmer interface to any number of solvers, linear algebra libraries and preconditioner packages, providing also a sophisticated technique for combining multiple packages to solve a single problem. TSF provides a collection of abstract base classes that define the interfaces to abstract vector, matrix and linear soeratormoreĀ Ā» objects. By using abstract interfaces, users of TSF are not limiting themselves to any one concrete library and can in fact easily combine multiple libraries to solve a single problem.Ā«Ā less

  20. Mixed waste: Proceedings

    SciTech Connect (OSTI)

    Moghissi, A.A.; Blauvelt, R.K.; Benda, G.A.; Rothermich, N.E.

    1993-12-31

    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.

  1. Tank Waste Committee Page 1

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

    ... Because of the stress on the tank due to it being bowed up in the middle (or "oil ... There would be no time for annulus pumping. The waste would almost instantly fill the leak ...

  2. Prototype heater test of the environment around a simulated waste...

    Office of Scientific and Technical Information (OSTI)

    Prototype heater test of the environment around a simulated waste package Citation Details In-Document Search Title: Prototype heater test of the environment around a simulated ...

  3. Cleanup Verification Package for the 618-3 Burial Ground

    SciTech Connect (OSTI)

    M. J. Appel

    2006-09-12

    This cleanup verification package documents completion of remedial action for the 618-3 Solid Waste Burial Ground, also referred to as Burial Ground Number 3 and the Dry Waste Burial Ground Number 3. During its period of operation, the 618-3 site was used to dispose of uranium-contaminated construction debris from the 311 Building and construction/demolition debris from remodeling of the 313, 303-J and 303-K Buildings.

  4. Automatic Differentiation Package

    Energy Science and Technology Software Center (OSTI)

    2007-03-01

    Sacado is an automatic differentiation package for C++ codes using operator overloading and C++ templating. Sacado provide forward, reverse, and Taylor polynomial automatic differentiation classes and utilities for incorporating these classes into C++ codes. Users can compute derivatives of computations arising in engineering and scientific applications, including nonlinear equation solving, time integration, sensitivity analysis, stability analysis, optimization and uncertainity quantification.

  5. Packaging and Transportation Safety

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

    2003-04-04

    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.

  6. Jpetra Kernel Package

    Energy Science and Technology Software Center (OSTI)

    2004-03-01

    A package of classes for constructing and using distributed sparse and dense matrices, vectors and graphs, written in Java. Jpetra is intended to provide the foundation for basic matrix and vector operations for Java developers. Jpetra provides distributed memory operations via an abstract parallel machine interface. The most common implementation of this interface will be Java sockets.

  7. Particle-filled microporous materials

    DOE Patents [OSTI]

    McAllister, J.W.; Kinzer, K.E.; Mrozinski, J.S.; Johnson, E.J.; Dyrud, J.F.

    1990-09-18

    A microporous particulate-filled thermoplastic polymeric article is provided. The article can be in the form of a film, a fiber, or a tube. The article has a thermoplastic polymeric structure having a plurality of interconnected passageways to provide a network of communicating pores. The microporous structure contains discrete submicron or low micron-sized particulate filler, the particulate filler being substantially non-agglomerated. 3 figs.

  8. Particle-filled microporous materials

    DOE Patents [OSTI]

    McAllister, Jerome W.; Kinzer, Kevin E.; Mrozinski, James S.; Johnson, Eric J.; Dyrud, James F.

    1990-01-01

    A microporous particulate-filled thermoplastic polymeric article is provided. The article can be in the form of a film, a fiber, or a tube. The article has a thermoplastic polymeric structure having a plurality of interconnected passageways to provide a network of communicating pores. The microporous structure contains discrete submicron or low micron-sized particulate filler, the particulate filler being substantially non-agglomerated.

  9. Particle-filled microporous materials

    DOE Patents [OSTI]

    McAllister, Jerome W.; Kinzer, Kevin E.; Mrozinski, James S.; Johnson, Eric J.

    1992-07-14

    A microporous particulate-filled thermoplastic polymeric article is provided. The article can be in the form of a film, a fiber, or a tube. The article has a thermoplastic polymeric structure having a plurality of interconnected passageways to provide a network of communicating pores. The microporous structure contains discrete submicron or low micron-sized particulate filler, the particulate filler being substantially non-agglomerated.

  10. High Efficiency Integrated Package

    SciTech Connect (OSTI)

    Ibbetson, James

    2013-09-15

    Solid-state lighting based on LEDs has emerged as a superior alternative to inefficient conventional lighting, particularly incandescent. LED lighting can lead to 80 percent energy savings; can last 50,000 hours ā€“ 2-50 times longer than most bulbs; and contains no toxic lead or mercury. However, to enable mass adoption, particularly at the consumer level, the cost of LED luminaires must be reduced by an order of magnitude while achieving superior efficiency, light quality and lifetime. To become viable, energy-efficient replacement solutions must deliver system efficacies of ā‰„ 100 lumens per watt (LPW) with excellent color rendering (CRI > 85) at a cost that enables payback cycles of two years or less for commercial applications. This development will enable significant site energy savings as it targets commercial and retail lighting applications that are most sensitive to the lifetime operating costs with their extended operating hours per day. If costs are reduced substantially, dramatic energy savings can be realized by replacing incandescent lighting in the residential market as well. In light of these challenges, Cree proposed to develop a multi-chip integrated LED package with an output of > 1000 lumens of warm white light operating at an efficacy of at least 128 LPW with a CRI > 85. This product will serve as the light engine for replacement lamps and luminaires. At the end of the proposed program, this integrated package was to be used in a proof-of-concept lamp prototype to demonstrate the componentā€™s viability in a common form factor. During this project Cree SBTC developed an efficient, compact warm-white LED package with an integrated remote color down-converter. Via a combination of intensive optical, electrical, and thermal optimization, a package design was obtained that met nearly all project goals. This package emitted 1295 lm under instant-on, room-temperature testing conditions, with an efficacy of 128.4 lm/W at a color temperature of ~2873K and 83 CRI. As such, the packageā€™s performance exceeds DOEā€™s warm-white phosphor LED efficacy target for 2013. At the end of the program, we assembled an A19 sized demonstration bulb housing the integrated package which met Energy Star intensity variation requirements. With further development to reduce overall component cost, we anticipate that an integrated remote converter package such as developed during this program will find application in compact, high-efficacy LED-based lamps, particularly those requiring omnidirectional emission.

  11. Using on-package memory

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

    on-package memory Using on-package memory Introduction The NERSC-8 system will include a novel feature on its node architecture: 16 GB of high-bandwidth 3D stacked memory...

  12. 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.

  13. Operational guidance for using DOT-6M/2R packaging

    SciTech Connect (OSTI)

    Kelly, D.L.; Hummer, J.H.

    1994-03-01

    The purpose of this paper is to describe a new US Department of Energy (DOE), Transportation Management Division task to create a US Department of Transportation (DOT) Specification 6M/2R packaging configuration user`s guide. The need for a user`s guide was identified because the DOT-6M/2R packaging configuration is widely used by DOE site contractors, and DOE receives many questions about the approved packaging configurations. Currently, two DOE organizations have the authority to approve new DOT-6M/2R configurations. For Defense Programs, the Transportation and Packaging Safety Division (EH-332) administers the program. For Environmental Restoration and Waste Management, the Transportation Management Division (EM-261) administers the program.

  14. Aquaculture information package

    SciTech Connect (OSTI)

    Boyd, T.; Rafferty, K.

    1998-08-01

    This package of information is intended to provide background information to developers of geothermal aquaculture projects. The material is divided into eight sections and includes information on market and price information for typical species, aquaculture water quality issues, typical species culture information, pond heat loss calculations, an aquaculture glossary, regional and university aquaculture offices and state aquaculture permit requirements. A bibliography containing 68 references is also included.

  15. Biodiesel Filling Stations UK | Open Energy Information

    Open Energy Info (EERE)

    Filling Stations UK Jump to: navigation, search Name: Biodiesel Filling Stations UK Place: United Kingdom Product: A website providing a list of places in the UK where people can...

  16. Plutonium stabilization and packaging system

    SciTech Connect (OSTI)

    1996-05-01

    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.

  17. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    SciTech Connect (OSTI)

    Schneider, K.J.; Andrews, W.B.; Schreiber, A.M.; Rosenthal, L.J.; Odle, C.J.

    1981-09-01

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes.

  18. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA

    SciTech Connect (OSTI)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION, NEVADA SITE OFFICE

    2005-07-01

    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. Packaging - Materials review

    SciTech Connect (OSTI)

    Herrmann, Matthias

    2014-06-16

    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. Aristos Optimization Package

    Energy Science and Technology Software Center (OSTI)

    2007-03-01

    Aristos is a Trilinos package for nonlinear continuous optimization, based on full-space sequential quadratic programming (SQP) methods. Aristos is specifically designed for the solution of large-scale constrained optimization problems in which the linearized constraint equations require iterative (i.e. inexact) linear solver techniques. Aristos' unique feature is an efficient handling of inexactness in linear system solves. Aristos currently supports the solution of equality-constrained convex and nonconvex optimization problems. It has been used successfully in the areamoreĀ Ā» of PDE-constrained optimization, for the solution of nonlinear optimal control, optimal design, and inverse problems.Ā«Ā less

  1. Laser diode package with enhanced cooling

    DOE Patents [OSTI]

    Deri, Robert J.; Kotovsky, Jack; Spadaccini, Christopher M.

    2011-09-13

    A laser diode package assembly includes a reservoir filled with a fusible metal in close proximity to a laser diode. The fusible metal absorbs heat from the laser diode and undergoes a phase change from solid to liquid during the operation of the laser. The metal absorbs heat during the phase transition. Once the laser diode is turned off, the liquid metal cools off and resolidifies. The reservoir is designed such that that the liquid metal does not leave the reservoir even when in liquid state. The laser diode assembly further includes a lid with one or more fin structures that extend into the reservoir and are in contact with the metal in the reservoir.

  2. Laser diode package with enhanced cooling

    DOE Patents [OSTI]

    Deri, Robert J.; Kotovsky, Jack; Spadaccini, Christopher M.

    2012-06-12

    A laser diode package assembly includes a reservoir filled with a fusible metal in close proximity to a laser diode. The fusible metal absorbs heat from the laser diode and undergoes a phase change from solid to liquid during the operation of the laser. The metal absorbs heat during the phase transition. Once the laser diode is turned off, the liquid metal cools off and resolidifies. The reservoir is designed such that that the liquid metal does not leave the reservoir even when in liquid state. The laser diode assembly further includes a lid with one or more fin structures that extend into the reservoir and are in contact with the metal in the reservoir.

  3. Laser diode package with enhanced cooling

    DOE Patents [OSTI]

    Deri, Robert J.; Kotovsky, Jack; Spadaccini, Christopher M.

    2012-06-26

    A laser diode package assembly includes a reservoir filled with a fusible metal in close proximity to a laser diode. The fusible metal absorbs heat from the laser diode and undergoes a phase change from solid to liquid during the operation of the laser. The metal absorbs heat during the phase transition. Once the laser diode is turned off, the liquid metal cools off and resolidifies. The reservoir is designed such that that the liquid metal does not leave the reservoir even when in liquid state. The laser diode assembly further includes a lid with one or more fin structures that extend into the reservoir and are in contact with the metal in the reservoir.

  4. Defect-enhanced void filling and novel filled phases of open-structure skutterudites

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Xi, Lili; Qiu, Yuting; Zhang, Wenqing; Chen, Lidong; Singh, David J.; Yang, Jihui

    2015-05-14

    Here, we report the design of novel filled CoSb3 skutterudite phases based on a combination of filling and Sb-substituted Ga/In defects. Ga/In doped skutterudite phases with Li-, Nd-, and Sm-fillings can be formed via this strategy, which can have relatively wider ranges of carrier concentration than other conventional filled skutterudite phases.

  5. Defect-enhanced void filling and novel filled phases of open-structure skutterudites

    SciTech Connect (OSTI)

    Xi, Lili; Qiu, Yuting; Zhang, Wenqing; Chen, Lidong; Singh, David J.; Yang, Jihui

    2015-05-14

    Here, we report the design of novel filled CoSb3 skutterudite phases based on a combination of filling and Sb-substituted Ga/In defects. Ga/In doped skutterudite phases with Li-, Nd-, and Sm-fillings can be formed via this strategy, which can have relatively wider ranges of carrier concentration than other conventional filled skutterudite phases.

  6. Experimental investigation on hydrogen cryogenic distillation equipped with package made by ICIT

    SciTech Connect (OSTI)

    Bornea, A.; Zamfirache, M.; Stefan, L.; Stefanescu, I.; Preda, A.

    2015-03-15

    ICIT (Institute for Cryogenics and Isotopic Technologies) has used its experience in cryogenic water distillation process to propose a similar process for hydrogen distillation that can be used in detritiation technologies. This process relies on the same packages but a stainless filling is tested instead of the phosphorous bronze filling used for water distillation. This paper presents two types of packages developed for hydrogen distillation, both have a stainless filling but it differs in terms of density, exchange surface and specific volume. Performance data have been obtained on laboratory scale. In order to determine the characteristics of the package, the installation was operated in the total reflux mode, for different flow rate for the liquid. There were made several experiments considering different operating conditions. Samples extracted at the top and bottom of cryogenic distillation column allowed mathematical processing to determine the separation performance. The experiments show a better efficiency for the package whose exchange surface was higher and there were no relevant differences between both packages as the operating pressure of the cryogenic column was increasing. For a complete characterization of the packages, future experiments will be considered to determine performance at various velocities in the column and their correlation with the pressure in the column. We plan further experiments to separate tritium from the mixture of isotopes DT, having in view that our goal is to apply this results to a detritiation plant.

  7. IDAHO SITE TO PROVIDE WASTE TREATMENT FOR OTHER DOE SITES

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

    March 7, 2008 IDAHO SITE TO PROVIDE WASTE TREATMENT FOR OTHER DOE SITES Plan won't impact DOE commitment to removing all stored waste from Idaho Site Idaho's Advanced Mixed Waste Treatment Facility offers state of the art waste characterization, treatment and packaging capabilities. Click on image to enlarge The U.S. Department of Energy (DOE) is amending the Record of Decision for the Waste Management Program: Treatment and Storage of Transuranic Waste, originally issued in 1998. The amendment

  8. Remaining Sites Verification Package for the 100-B-21:2 Subsite (100-B/C Discovery Pipeline DS-100BC-002), Waste Site Reclassification Form 2008-003

    SciTech Connect (OSTI)

    J. M. Capron

    2008-06-16

    The 100-B-21:2 waste site consists of the immediate area of the DS-100BC-02 pipeline. In accordance with this evaluation, the confirmatory and 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.

  9. Transmittal of the Calculation Package that Supports the Analysis of Performance of the Environmental Management Waste Management Facility Oak Ridge, Tennessee (Based 5-Cell Design Issued 8/14/09)

    SciTech Connect (OSTI)

    Williams M.J.

    2009-09-14

    This document presents the results of an assessment of the performance of a build-out of the Environmental Management Waste Management Facility (EMWMF). The EMWMF configuration that was assessed includes the as-constructed Cells 1 through 4, with a groundwater underdrain that was installed beneath Cell 3 during the winter of 2003-2004, and Cell 5, whose proposed design is an Addendum to Remedial Design Report for the Disposal of Oak Ridge Reservation Comprehensive Environmental Response, Compensation, and Liability Act of 1980 Waste, Oak Ridge, Tennessee, DOE/OR/01-1873&D2/A5/R1. The total capacity of the EMWMF with 5 cells is about 1.7 million cubic yards. This assessment was conducted to determine the conditions under which the approved Waste Acceptance Criteria (WAC) for the EMWMF found in the Attainment Plan for Risk/Toxicity-Based Waste Acceptance Criteria at the Oak Ridge Reservation, Oak Ridge, Tennessee [U.S. Department of Energy (DOE) 2001a], as revised for constituents added up to October 2008, would remain protective of public health and safety for a five-cell disposal facility. For consistency, the methods of analyses and the exposure scenario used to predict the performance of a five-cell disposal facility were identical to those used in the Remedial Investigation and Feasibility Study (RI/FS) and its addendum (DOE 1998a, DOE 1998b) to develop the approved WAC. To take advantage of new information and design changes departing from the conceptual design, the modeling domain and model calibration were upaded from those used in the RI/FS and its addendum. It should be noted that this analysis is not intended to justify or propose a change in the approved WAC.

  10. Consolidation process for producing ceramic waste forms

    DOE Patents [OSTI]

    Hash, Harry C.; Hash, Mark C.

    2000-01-01

    A process for the consolidation and containment of solid or semisolid hazardous waste, which process comprises closing an end of a circular hollow cylinder, filling the cylinder with the hazardous waste, and then cold working the cylinder to reduce its diameter while simultaneously compacting the waste. The open end of the cylinder can be sealed prior to or after the cold working process. The preferred method of cold working is to draw the sealed cylinder containing the hazardous waste through a plurality of dies to simultaneously reduce the diameter of the tube while compacting the waste. This process provides a quick continuous process for consolidating hazardous waste, including radioactive waste.

  11. Hanford site transuranic waste certification plan

    SciTech Connect (OSTI)

    GREAGER, T.M.

    1999-05-12

    As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of U.S. Department of Energy (DOE) Order 5820.2A, ''Radioactive Waste Management, and the Waste Acceptance Criteria for the Waste Isolation Pilot Plant' (DOE 1996d) (WIPP WAC). The WIPP WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WIPP WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their management of TRU waste and TRU waste shipments before transferring waste to WIPP. The Hanford Site must also ensure that its TRU waste destined for disposal at WIPP meets requirements for transport in the Transuranic Package Transporter41 (TRUPACT-11). The U.S. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-I1 requirements in the ''Safety Analysis Report for the TRUPACT-II Shipping Package'' (NRC 1997) (TRUPACT-I1 SARP).

  12. Thyra Abstract Interface Package

    Energy Science and Technology Software Center (OSTI)

    2005-09-01

    Thrya primarily defines a set of abstract C++ class interfaces needed for the development of abstract numerical atgorithms (ANAs) such as iterative linear solvers, transient solvers all the way up to optimization. At the foundation of these interfaces are abstract C++ classes for vectors, vector spaces, linear operators and multi-vectors. Also included in the Thyra package is C++ code for creating concrete vector, vector space, linear operator, and multi-vector subclasses as well as other utilitiesmoreĀ Ā» to aid in the development of ANAs. Currently, very general and efficient concrete subclass implementations exist for serial and SPMD in-core vectors and multi-vectors. Code also currently exists for testing objects and providing composite objects such as product vectors.Ā«Ā less

  13. Meros Preconditioner Package

    Energy Science and Technology Software Center (OSTI)

    2004-04-01

    Meros uses the compositional, aggregation, and overload operator capabilities of TSF to provide an object-oriented package providing segregated/block preconditioners for linear systems related to fully-coupled Navier-Stokes problems. This class of preconditioners exploits the special properties of these problems to segregate the equations and use multi-level preconditioners (through ML) on the matrix sub-blocks. Several preconditioners are provided, including the Fp and BFB preconditioners of Kay & Loghin and Silvester, Elman, Kay & Wathen. The overall performancemoreĀ Ā» and scalability of these preconditioners approaches that of multigrid for certain types of problems. Meros also provides more traditional pressure projection methods including SIMPLE and SIMPLEC.Ā«Ā less

  14. Tpetra Kernel Package

    Energy Science and Technology Software Center (OSTI)

    2004-03-01

    A package of classes for constructing and using distributed sparse and dense matrices, vectors and graphs. Templated on the scalar and ordinal types so that any valid floating-point type, as well as any valid integer type can be used with these classes. Other non-standard types, such as 3-by-3 matrices for the scalar type and mod-based integers for ordinal types, can also be used. Tpetra is intended to provide the foundation for basic matrix and vectormoreĀ Ā» operations for the next generation of Trilinos preconditioners and solvers, It can be considered as the follow-on to Epetra. Tpetra provides distributed memory operations via an abstract parallel machine interface, The most common implementation of this interface will be MPI.Ā«Ā less

  15. Anasazi Block Eigensolvers Package

    Energy Science and Technology Software Center (OSTI)

    2004-03-01

    ANASAZI is an extensible and interoperable framework for large-scale eigenvalue algorithms. The motivation for this framework is to provide a generic interface to a collection of algorithms for solving large-scale eigenvalue problems. ANASAZI is interoperable because both the matrix and vectors (defining the eigenspace) are considered to be opaque objects---only knowledge of the matrix and vectors via elementary operations is necessary. An implementation of Anasazi is accomplished via the use of interfaces. One of themoreĀ Ā» goals of ANASAZI is to allow the user the flexibility to specify the data representation for the matrix and vectors and so leverage any existing software investment. The algorithms that will be included in package are Krylov-based and preconditioned eigensolvers.Ā«Ā less

  16. Piecewise Cubic Interpolation Package

    Energy Science and Technology Software Center (OSTI)

    1982-04-23

    PCHIP (Piecewise Cubic Interpolation Package) is a set of subroutines for piecewise cubic Hermite interpolation of data. It features software to produce a monotone and "visually pleasing" interpolant to monotone data. Such an interpolant may be more reasonable than a cubic spline if the data contain both 'steep' and 'flat' sections. Interpolation of cumulative probability distribution functions is another application. In PCHIP, all piecewise cubic functions are represented in cubic Hermite form; that is, f(x)moreĀ Ā» is determined by its values f(i) and derivatives d(i) at the breakpoints x(i), i=1(1)N. PCHIP contains three routines - PCHIM, PCHIC, and PCHSP to determine derivative values, six routines - CHFEV, PCHFE, CHFDV, PCHFD, PCHID, and PCHIA to evaluate, differentiate, or integrate the resulting cubic Hermite function, and one routine to check for monotonicity. A FORTRAN 77 version and SLATEC version of PCHIP are included.Ā«Ā less

  17. Electro-Microfluidic Packaging

    SciTech Connect (OSTI)

    BENAVIDES, GILBERT L.; GALAMBOS, PAUL C.

    2002-06-01

    Electro-microfluidics is experiencing explosive growth in new product developments. There are many commercial applications for electro-microfluidic devices such as chemical sensors, biological sensors, and drop ejectors for both printing and chemical analysis. The number of silicon surface micromachined electro-microfluidic products is likely to increase. Manufacturing efficiency and integration of microfluidics with electronics will become important. Surface micromachined microfluidic devices are manufactured with the same tools as IC's (integrated circuits) and their fabrication can be incorporated into the IC fabrication process. In order to realize applications for devices must be developed. An Electro-Microfluidic Dual In-line Package (EMDIP{trademark}) was developed surface micromachined electro-microfluidic devices, a practical method for getting fluid into these to be a standard solution that allows for both the electrical and the fluidic connections needed to operate a great variety of electro-microfluidic devices. The EMDIP{trademark} includes a fan-out manifold that, on one side, mates directly with the 200 micron diameter Bosch etched holes found on the device, and, on the other side, mates to lager 1 mm diameter holes. To minimize cost the EMDIP{trademark} can be injection molded in a great variety of thermoplastics which also serve to optimize fluid compatibility. The EMDIP{trademark} plugs directly into a fluidic printed wiring board using a standard dual in-line package pattern for the electrical connections and having a grid of multiple 1 mm diameter fluidic connections to mate to the underside of the EMDIP{trademark}.

  18. Remaining Sites Verification Package for the 100-F-26:10, 1607-F3 Sanitary Sewer Pipelines (182-F, 183-F, and 151-F Sanitary Sewer Lines), Waste Site Reclassification Form 2007-028

    SciTech Connect (OSTI)

    L. M. Dittmer

    2007-12-03

    The 100-F-26:10 waste site includes sanitary sewer lines that serviced the former 182-F, 183-F, and 151-F Buildings. 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.

  19. Characterization of Carbon Epoxy-Filled Composite

    SciTech Connect (OSTI)

    Mason, Thomas Albert; Liu, Cheng; Lovato, Manuel L.; Valdez, James Anthony; Cady, Carl Mcelhinney; Walker, Emily Kristine; Livescu, Veronica

    2015-11-25

    Please find attached a summary of the characterization work performed at Los Alamos between 2014 and 2015 on epoxy-filled carbon composite material.

  20. 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

    2013-07-01

    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)

  1. Bulk Tritium Shipping Package | Department of Energy

    Office of Environmental Management (EM)

    Package Bulk Tritium Shipping Package Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013. PDF icon Bulk Tritium Shipping Package More Documents & Publications Bulk Tritium Shipping Package Overview and Status Managing Legacy Materials at WETF FAQS Reference Guide - NNSA Package Certification Engineer

  2. Nevada Test Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2005-10-01

    This document establishes the U.S. Department of Energy (DOE), 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 (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) 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 NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  3. Low-level waste program technical strategy

    SciTech Connect (OSTI)

    Bledsoe, K.W.

    1994-10-01

    The Low-Level Waste Technical Strategy document describes the mechanisms which the Low-Level Waste Program Office plans to implement to achieve its mission. The mission is to manage the receipt, immobilization, packaging, storage/disposal and RCRA closure (of the site) of the low-level Hanford waste (pretreated tank wastes) in an environmentally sound, safe and cost-effective manner. The primary objective of the TWRS Low-level waste Program office is to vitrify the LLW fraction of the tank waste and dispose of it onsite.

  4. Packaging, Transportation and Recycling of NPP Condenser Modules - 12262

    SciTech Connect (OSTI)

    Polley, G.M. [Perma-Fix Environmental Services, 575 Oak Ridge Turnpike, Oak Ridge, TN 37830 (United States)

    2012-07-01

    Perma-Fix was awarded contract from Energy Northwest for the packaging, transportation and disposition of the condenser modules, water boxes and miscellaneous metal, combustibles and water generated during the 2011 condenser replacement outage at the Columbia Generating Station. The work scope was to package the water boxes and condenser modules as they were removed from the facility and transfer them to the Perma-Fix Northwest facility for processing, recycle of metals and disposition. The condenser components were oversized and overweight (the condenser modules weighed ?102,058 kg [225,000 lb]) which required special equipment for loading and transport. Additional debris waste was packaged in inter-modals and IP-1 boxes for transport. A waste management plan was developed to minimize the generation of virtually any waste requiring landfill disposal. The Perma-Fix Northwest facility was modified to accommodate the ?15 m [50-ft] long condenser modules and equipment was designed and manufactured to complete the disassembly, decontamination and release survey. The condenser modules are currently undergoing processing for free release to a local metal recycler. Over three millions pounds of metal will be recycled and over 95% of the waste generated during this outage will not require land disposal. There were several elements of this project that needed to be addressed during the preparation for this outage and the subsequent packaging, transportation and processing. - Staffing the project to support 24/7 generation of large components and other wastes. - The design and manufacture of the soft-sided shipping containers for the condenser modules that measured ?15 m X 4 m X 3 m [50 ft X 13 ft X 10 ft] and weighed ?102,058 kg [225,000 lbs] - Developing a methodology for loading the modules into the shipping containers. - Obtaining a transport vehicle for the modules. - Designing and modifying the processing facility. - Movement of the modules at the processing facility. If any of these issues were not adequately resolved prior to the start of the outage, costly delays would result and the re-start of the power plant could be impacted. The main focus of this project was to find successful methods for keeping this material out of the landfills and preserving the natural resources. In addition, this operation provided a significant cost savings to the public utility by minimizing landfill disposal. The onsite portion of the project has been completed without impact to the overall outage schedule. By the date of presentation, the majority of the waste from the condenser replacement project will have been processed and recycled. The goals for this project included helping Energy Northwest maintain the outage schedule, package and characterize waste compliantly, perform transportation activities in compliance with 49CFR (Ref-1), and minimize the waste disposal volume. During this condenser replacement project, over three millions pounds of waste was generated, packaged, characterized and transported without injury or incident. It is anticipated that 95% of the waste generated during this project will not require landfill disposal. All of the waste is scheduled to be processed, decontaminated and recycled by June of 2012. (authors)

  5. Amesos Solver Package

    Energy Science and Technology Software Center (OSTI)

    2004-03-01

    Amesos is the Direct Sparse Solver Package in Trilinos. The goal of Amesos is to make AX=S as easy as it sounds, at least for direct methods. Amesos provides interfaces to a number of third party sparse direct solvers, including SuperLU, SuperLU MPI, DSCPACK, UMFPACK and KLU. Amesos provides a common object oriented interface to the best sparse direct solvers in the world. A sparse direct solver solves for x in Ax = b. wheremoreĀ Ā» A is a matrix and x and b are vectors (or multi-vectors). A sparse direct solver flrst factors A into trinagular matrices L and U such that A = LU via gaussian elimination and then solves LU x = b. Switching amongst solvers in Amesos roquires a change to a single parameter. Yet, no solver needs to be linked it, unless it is used. All conversions between the matrices provided by the user and the format required by the underlying solver is performed by Amesos. As new sparse direct solvers are created, they will be incorporated into Amesos, allowing the user to simpty link with the new solver, change a single parameter in the calling sequence, and use the new solver. Amesos allows users to specify whether the matrix has changed. Amesos can be used anywhere that any sparse direct solver is needed.Ā«Ā less

  6. Part IV: Section D - Packaging and Marking

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

    PART I SECTION D PACKAGING AND MARKING DE-AC36-08GO28308 Modification M901 Section D - Page ii PART I SECTION D PACKAGING AND MARKING TABLE OF CONTENTS D.1 Packaging 1 D.2 Marking ...

  7. Power Device Packaging | Department of Energy

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

    D.C. PDF icon ape023wang2010p.pdf More Documents & Publications Power Device Packaging Power Device Packaging Direct Water-Cooled Power Electronics Substrate Packaging

  8. Using on-package memory

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

    on-package memory Using on-package memory Introduction The NERSC-8 system will include a novel feature on its node architecture: 16 GB of high-bandwidth 3D stacked memory interposed between the KNL chip and the slower off-package DDR memory. Compared to the on-node DDR4 memory, the high-bandwidth memory (HBM) has approximately 5x the bandwidth but has similar latency. This new feature has the potential to accelerate those applications which are particularly sensitive to memory bandwidth limits.

  9. Packaging of solid state devices

    DOE Patents [OSTI]

    Glidden, Steven C.; Sanders, Howard D.

    2006-01-03

    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.

  10. Safety evaluation for packaging CPC metal boxes

    SciTech Connect (OSTI)

    Romano, T.

    1995-05-15

    This Safety Evaluation for Packaging (SEP) provides authorization for the use of Container Products Corporation (CPC) metal boxes, as described in this document, for the interarea shipment of radioactive contaminated equipment and debris for storage in the Central Waste Complex (CWC) or T Plant located in the 200 West Area. Authorization is granted until November 30, 1995. The CPC boxes included in this SEP were originally procured as US Department of Transportation (DOT) Specification 7A Type A boxes. A review of the documentation provided by the manufacturer revealed the documentation did not adequately demonstrate compliance to the 4 ft drop test requirement of 49 CFR 173.465(c). Preparation of a SEP is necessary to document the equivalent safety of the onsite shipment in lieu of meeting DOT packaging requirements until adequate documentation is received. The equivalent safety of the shipment is based on the fact that the radioactive contents consist of contaminated equipment and debris which are not dispersible. Each piece is wrapped in two layers of no less than 4 mil plastic prior to being placed in the box which has an additional 10 mil liner. Pointed objects and sharp edges are padded to prevent puncture of the plastic liner and wrapping.

  11. Hanford Site Transuranic (TRU) Waste Certification Plan

    SciTech Connect (OSTI)

    GREAGER, T.M.

    2000-12-01

    As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of US. Department of Energy (DOE) 0 435.1, ''Radioactive Waste Management,'' and the Contact-Handled (CH) Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC). WIPP-WAC requirements are derived from the WIPP Technical Safety Requirements, WIPP Safety Analysis Report, TRUPACT-II SARP, WIPP Land Withdrawal Act, WIPP Hazardous Waste Facility Permit, and Title 40 Code of Federal Regulations (CFR) 191/194 Compliance Certification Decision. The WIPP-WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WPP-WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their program for managing TRU waste and TRU waste shipments before transferring waste to WIPP. Waste characterization activities provide much of the data upon which certification decisions are based. Waste characterization requirements for TRU waste and TRU mixed waste that contains constituents regulated under the Resource Conservation and Recovery Act (RCRA) are established in the WIPP Hazardous Waste Facility Permit Waste Analysis Plan (WAP). The Hanford Site Quality Assurance Project Plan (QAPjP) (HNF-2599) implements the applicable requirements in the WAP and includes the qualitative and quantitative criteria for making hazardous waste determinations. The Hanford Site must also ensure that its TRU waste destined for disposal at WPP meets requirements for transport in the Transuranic Package Transporter-11 (TRUPACT-11). The US. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-11 requirements in the Safety Analysis Report for the TRUPACT-II Shipping Package (TRUPACT-11 SARP). In addition, a TRU waste is eligible for disposal at WIPP only if it has been generated in whole or in part by one or more of the activities listed in Section 10101(3) of the Nuclear Waste Policy Act. DOE sites must determine that each waste stream to be disposed of at WIPP is ''defense'' TRU waste. (See also the definition of ''defense'' TRU waste.). Only CH TRU wastes meeting the requirements of the QAPjP, WIPP-WAP, WPP-WAC, and other requirements documents described above will be accepted for transportation and disposal at WIPP.

  12. Hanford Site Transuranic (TRU) Waste Certification Plan

    SciTech Connect (OSTI)

    GREAGER, T.M.

    2000-12-06

    As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of US. Department of Energy (DOE) 0 435.1, ''Radioactive Waste Management,'' and the Contact-Handled (CH) Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC). WIPP-WAC requirements are derived from the WIPP Technical Safety Requirements, WIPP Safety Analysis Report, TRUPACT-II SARP, WIPP Land Withdrawal Act, WIPP Hazardous Waste Facility Permit, and Title 40 Code of Federal Regulations (CFR) 191/194 Compliance Certification Decision. The WIPP-WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WPP-WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their program for managing TRU waste and TRU waste shipments before transferring waste to WIPP. Waste characterization activities provide much of the data upon which certification decisions are based. Waste characterization requirements for TRU waste and TRU mixed waste that contains constituents regulated under the Resource Conservation and Recovery Act (RCRA) are established in the WIPP Hazardous Waste Facility Permit Waste Analysis Plan (WAP). The Hanford Site Quality Assurance Project Plan (QAPjP) (HNF-2599) implements the applicable requirements in the WAP and includes the qualitative and quantitative criteria for making hazardous waste determinations. The Hanford Site must also ensure that its TRU waste destined for disposal at WPP meets requirements for transport in the Transuranic Package Transporter-11 (TRUPACT-11). The US. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-11 requirements in the Safety Analysis Report for the TRUPACT-II Shipping Package (TRUPACT-11 SARP). In addition, a TRU waste is eligible for disposal at WIPP only if it has been generated in whole or in part by one or more of the activities listed in Section 10101(3) of the Nuclear Waste Policy Act. DOE sites must determine that each waste stream to be disposed of at WIPP is ''defense'' TRU waste. (See also the definition of ''defense'' TRU waste.). Only CH TRU wastes meeting the requirements of the QAPjP, WIPP-WAP, WPP-WAC, and other requirements documents described above will be accepted for transportation and disposal at WIPP.

  13. Packaging and Transportation News | Department of Energy

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

    Packaging and Transportation News January 14, 2016 Ron Hafner with Lawrence Livermore National Laboratory lectures for a course in San Ramon, Calif. on packaging and transporting ...

  14. Review of SAR for Packaging Report

    Broader source: Energy.gov [DOE]

    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.

  15. Handling and Packaging a Potentially Radiologically Contaminated...

    Office of Environmental Management (EM)

    Handling and Packaging a Potentially Radiologically Contaminated Patient Handling and Packaging a Potentially Radiologically Contaminated Patient The purpose of this procedure is...

  16. Hazardous Materials Packaging and Transportation Safety - DOE...

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

    60.1D, Hazardous Materials Packaging and Transportation Safety by Ashok Kapoor Functional areas: Hazardous Materials, Packaging and Transportation, Safety and Security, Work...

  17. Building America Webinar: Standardized Retrofit Packages - What...

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

    Standardized Retrofit Packages - What Works to Meet Consistent Levels of Performance? Building America Webinar: Standardized Retrofit Packages - What Works to Meet Consistent ...

  18. Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

    SciTech Connect (OSTI)

    Laurinat, James E.; Askew, Neal M.; Hensel, Steve J.

    2013-03-21

    Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

  19. 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

    2007-04-12

    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.

  20. Hanford Site Secondary Waste Roadmap

    SciTech Connect (OSTI)

    Westsik, Joseph H.

    2009-01-29

    Summary The U.S. Department of Energy (DOE) is making plans to dispose of 54 million gallons of radioactive tank wastes at the Hanford Site near Richland, Washington. The high-level wastes and low-activity wastes will be vitrified and placed in permanent disposal sites. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents, and these need to be processed and disposed of also. The Department of Energy Office of Waste Processing sponsored a meeting to develop a roadmap to outline the steps necessary to design the secondary waste forms. Representatives from DOE, the U.S. Environmental Protection Agency, the Washington State Department of Ecology, the Oregon Department of Energy, Nuclear Regulatory Commission, technical experts from the DOE national laboratories, academia, and private consultants convened in Richland, Washington, during the week of July 21-23, 2008, to participate in a workshop to identify the risks and uncertainties associated with the treatment and disposal of the secondary wastes and to develop a roadmap for addressing those risks and uncertainties. This report describes the results of the roadmap meeting in Richland. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents. The secondary waste roadmap workshop focused on the waste streams that contained the largest fractions of the 129I and 99Tc that the Integrated Disposal Facility risk assessment analyses were showing to have the largest contribution to the estimated IDF disposal impacts to groundwater. Thus, the roadmapping effort was to focus on the scrubber/off-gas treatment liquids with 99Tc to be sent to the Effluent Treatment Facility for treatment and solidification and the silver mordenite and carbon beds with the captured 129I to be packaged and sent to the IDF. At the highest level, the secondary waste roadmap includes elements addressing regulatory and performance requirements, waste composition, preliminary waste form screening, waste form development, process design and support, and validation. The regulatory and performance requirements activity will provide the secondary waste-form performance requirements. The waste-composition activity will provide workable ranges of secondary waste compositions and formulations for simulants and surrogates. Preliminary waste form screening will identify candidate waste forms for immobilizing the secondary wastes. The waste form development activity will mature the waste forms, leading to a selected waste form(s) with a defensible understanding of the long-term release rate and input into the critical decision process for a secondary waste treatment process/facility. The process and design support activity will provide a reliable process flowsheet and input to support a robust facility design. The validation effort will confirm that the selected waste form meets regulatory requirements. The final outcome of the implementation of the secondary waste roadmap is the compliant, effective, timely, and cost-effective disposal of the secondary wastes. The work necessary to address the programmatic, regulatory, and technical risks and uncertainties identified through the Secondary Waste Roadmap Workshop are assembled into several program needs elements. Programmatic/Regulatory needs include: ā€¢ Select and deploy Hanford tank waste supplemental treatment technology ā€¢ Provide treatment capability for secondary waste streams from tank waste treatment ā€¢ Develop consensus on secondary waste form acceptance. Technology needs include: ā€¢ Define secondary waste composition ranges and uncertainties ā€¢ Identify and develop waste forms for secondary waste immobilization and disposal ā€¢ Develop test methods to characterize secondary waste form performance. Details for each of these program elements are provided.

  1. Alternative Fuels Data Center: Filling CNG Fuel Tanks

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    Filling CNG Fuel Tanks to someone by E-mail Share Alternative Fuels Data Center: Filling CNG Fuel Tanks on Facebook Tweet about Alternative Fuels Data Center: Filling CNG Fuel Tanks on Twitter Bookmark Alternative Fuels Data Center: Filling CNG Fuel Tanks on Google Bookmark Alternative Fuels Data Center: Filling CNG Fuel Tanks on Delicious Rank Alternative Fuels Data Center: Filling CNG Fuel Tanks on Digg Find More places to share Alternative Fuels Data Center: Filling CNG Fuel Tanks on

  2. TANK 18-F AND 19-F TANK FILL GROUT SCALE UP TEST SUMMARY

    SciTech Connect (OSTI)

    Stefanko, D.; Langton, C.

    2012-01-03

    High-level waste (HLW) tanks 18-F and 19-F have been isolated from FTF facilities. To complete operational closure the tanks will be filled with grout for the purpose of: (1) physically stabilizing the tanks, (2) limiting/eliminating vertical pathways to residual waste, (3) entombing waste removal equipment, (4) discouraging future intrusion, and (5) providing an alkaline, chemical reducing environment within the closure boundary to control speciation and solubility of select radionuclides. This report documents the results of a four cubic yard bulk fill scale up test on the grout formulation recommended for filling Tanks 18-F and 19-F. Details of the scale up test are provided in a Test Plan. The work was authorized under a Technical Task Request (TTR), HLE-TTR-2011-008, and was performed according to Task Technical and Quality Assurance Plan (TTQAP), SRNL-RP-2011-00587. The bulk fill scale up test described in this report was intended to demonstrate proportioning, mixing, and transportation, of material produced in a full scale ready mix concrete batch plant. In addition, the material produced for the scale up test was characterized with respect to fresh properties, thermal properties, and compressive strength as a function of curing time.

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

    SciTech Connect (OSTI)

    Sanchez, Mike; Reeves, Wendall; Smart, Bill

    2013-07-01

    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)

  4. NREL: Transportation Research - Power Electronics Packaging Reliability

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

    Packaging Reliability A photo of a piece of power electronics testing equipment. NREL power electronics packaging reliability research investigates the performance and reliability of emerging interconnection, interface, and packaging materials. Findings help improve reliability and durability of emerging technologies. Photo by Dennis Schroeder, NREL Power electronics packaging around a semiconductor switching device determines the electrical, thermal, and mechanical properties of a power

  5. Seal welded cast iron nuclear waste container

    DOE Patents [OSTI]

    Filippi, Arthur M.; Sprecace, Richard P.

    1987-01-01

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

  6. Secondary Waste Form Down-Selection Data Packageā€”Fluidized Bed Steam Reforming Waste Form

    SciTech Connect (OSTI)

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

    2011-09-12

    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.

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

    SciTech Connect (OSTI)

    Taylor, L.L.; Shikashio, R.

    1993-09-01

    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.

  8. An Arbitrary Precision Computation Package

    Energy Science and Technology Software Center (OSTI)

    2003-06-14

    This package permits a scientist to perform computations using an arbitrarily high level of numeric precision (the equivalent of hundreds or even thousands of digits), by making only minor changes to conventional C++ or Fortran-90 soruce code. This software takes advantage of certain properties of IEEE floating-point arithmetic, together with advanced numeric algorithms, custom data types and operator overloading. Also included in this package is the "Experimental Mathematician's Toolkit", which incorporates many of these facilitiesmoreĀ Ā» into an easy-to-use interactive program.Ā«Ā less

  9. Process for remediation of plastic waste

    DOE Patents [OSTI]

    Pol, Vilas G; Thiyagarajan, Pappannan

    2013-11-12

    A single step process for degrading plastic waste by converting the plastic waste into carbonaceous products via thermal decomposition of the plastic waste by placing the plastic waste into a reactor, heating the plastic waste under an inert or air atmosphere until the temperature of about 700.degree. C. is achieved, allowing the reactor to cool down, and recovering the resulting decomposition products therefrom. The decomposition products that this process yields are carbonaceous materials, and more specifically carbon nanotubes having a partially filled core (encapsulated) adjacent to one end of the nanotube. Additionally, in the presence of a transition metal compound, this thermal decomposition process produces multi-walled carbon nanotubes.

  10. The First Recovery Act Funded Waste Shipment depart from the Advanced Mixed

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

    Waste Treatment Facility The First Recovery Act Funded Waste Shipment departs from the Advanced Mixed Waste Treatment Facility A shipment of mixed low-level waste left DOEļæ½s Advanced Mixed Waste Treatment Project on May 6, years earlier than originally planned, due to funding received through the American Recovery and Reinvestment Act. The 36.5 cubic meter shipment consisted of radioactive and chemically contaminated waste that had been retrieved, characterized, packaged and shipped from

  11. Package for integrated optic circuit and method

    DOE Patents [OSTI]

    Kravitz, S.H.; Hadley, G.R.; Warren, M.E.; Carson, R.F.; Armendariz, M.G.

    1998-08-04

    A structure and method are disclosed for packaging an integrated optic circuit. The package comprises a first wall having a plurality of microlenses formed therein to establish channels of optical communication with an integrated optic circuit within the package. A first registration pattern is provided on an inside surface of one of the walls of the package for alignment and attachment of the integrated optic circuit. The package in one embodiment may further comprise a fiber holder for aligning and attaching a plurality of optical fibers to the package and extending the channels of optical communication to the fibers outside the package. In another embodiment, a fiber holder may be used to hold the fibers and align the fibers to the package. The fiber holder may be detachably connected to the package. 6 figs.

  12. Package for integrated optic circuit and method

    DOE Patents [OSTI]

    Kravitz, Stanley H.; Hadley, G. Ronald; Warren, Mial E.; Carson, Richard F.; Armendariz, Marcelino G.

    1998-01-01

    A structure and method for packaging an integrated optic circuit. The package comprises a first wall having a plurality of microlenses formed therein to establish channels of optical communication with an integrated optic circuit within the package. A first registration pattern is provided on an inside surface of one of the walls of the package for alignment and attachment of the integrated optic circuit. The package in one embodiment may further comprise a fiber holder for aligning and attaching a plurality of optical fibers to the package and extending the channels of optical communication to the fibers outside the package. In another embodiment, a fiber holder may be used to hold the fibers and align the fibers to the package. The fiber holder may be detachably connected to the package.

  13. Nevada National Security Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2010-09-03

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

  14. Nevada National Security Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2011-01-01

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

  15. Radiation-induced mechanical property changes in filled rubber...

    Office of Scientific and Technical Information (OSTI)

    Radiation-induced mechanical property changes in filled rubber Citation Details In-Document Search Title: Radiation-induced mechanical property changes in filled rubber You are ...

  16. Radiation-induced mechanical property changes in filled rubber...

    Office of Scientific and Technical Information (OSTI)

    Radiation-induced mechanical property changes in filled rubber Citation Details In-Document Search Title: Radiation-induced mechanical property changes in filled rubber Authors: ...

  17. Multifunctional, Inorganic-Filled Separators for Large Format...

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

    & Publications Multifunctional, Inorganic-Filled Separators for Large Format, Li-ion Batteries Multifunctional, Inorganic-Filled Separators for Large Format, Li-ion Batteries...

  18. Mullins effect in a filled elastomer under uniaxial tension ...

    Office of Scientific and Technical Information (OSTI)

    Mullins effect in a filled elastomer under uniaxial tension Citation Details In-Document Search Title: Mullins effect in a filled elastomer under uniaxial tension Modulus softening ...

  19. EUHYFIS Hydrogen Filling Station Consortium | Open Energy Information

    Open Energy Info (EERE)

    EUHYFIS Hydrogen Filling Station Consortium Jump to: navigation, search Name: EUHYFIS (Hydrogen Filling Station Consortium) Place: Oldenburg, Germany Zip: 26123 Sector: Hydro,...

  20. Acceptable knowledge document for INEEL stored transuranic waste -- Rocky Flats Plant waste. Revision 2

    SciTech Connect (OSTI)

    1998-01-23

    This document and supporting documentation provide a consistent, defensible, and auditable record of acceptable knowledge for waste generated at the Rocky Flats Plant which is currently in the accessible storage inventory at the Idaho National Engineering and Environmental Laboratory. The inventory consists of transuranic (TRU) waste generated from 1972 through 1989. Regulations authorize waste generators and treatment, storage, and disposal facilities to use acceptable knowledge in appropriate circumstances to make hazardous waste determinations. Acceptable knowledge includes information relating to plant history, process operations, and waste management, in addition to waste-specific data generated prior to the effective date of the RCRA regulations. This document is organized to provide the reader a comprehensive presentation of the TRU waste inventory ranging from descriptions of the historical plant operations that generated and managed the waste to specific information about the composition of each waste group. Section 2 lists the requirements that dictate and direct TRU waste characterization and authorize the use of the acceptable knowledge approach. In addition to defining the TRU waste inventory, Section 3 summarizes the historical operations, waste management, characterization, and certification activities associated with the inventory. Sections 5.0 through 26.0 describe the waste groups in the inventory including waste generation, waste packaging, and waste characterization. This document includes an expanded discussion for each waste group of potential radionuclide contaminants, in addition to other physical properties and interferences that could potentially impact radioassay systems.

  1. EM Contractor's Success in Addressing Challenging Waste Leads...

    Office of Environmental Management (EM)

    Idaho Site cleanup contractor CH2M-WG Idaho's (CWI) success last year in treating and packaging 6,000 drums of radioactive and hazardous waste left over from the Cold War weapons...

  2. Alkaline earth filled nickel skutterudite antimonide thermoelectrics

    DOE Patents [OSTI]

    Singh, David Joseph

    2013-07-16

    A thermoelectric material including a body centered cubic filled skutterudite having the formula A.sub.xFe.sub.yNi.sub.zSb.sub.12, where A is an alkaline earth element, x is no more than approximately 1.0, and the sum of y and z is approximately equal to 4.0. The alkaline earth element includes guest atoms selected from the group consisting of Be, Mb, Ca, Sr, Ba, Ra and combinations thereof. The filled skutterudite is shown to have properties suitable for a wide variety of thermoelectric applications.

  3. DEVELOPMENT OF A NEW TYPE A(F)RADIOACTIVE MATERIAL PACKAGING FOR THE DEPARTMENT OF ENERGY

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2008-09-14

    In a coordinated effort, the Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) proposed the elimination of the Specification Packaging from 49 CFR 173.[1] In accordance with the Federal Register, issued on October 1, 2004, new fabrication of Specification Packages would no longer be authorized. In accordance with the NRC final rulemaking published January 26, 2004, Specification Packagings are mandated by law to be removed from service no later than October 1, 2008. This coordinated effort and resulting rulemaking initiated a planned phase out of Specification Type B and Type A fissile (F) material transportation packages within the Department of Energy (DOE) and its subcontractors. One of the Specification Packages affected by this regulatory change is the UN1A2 Specification Package, per DOT 49 CFR 173.417(a)(6). To maintain continuing shipments of DOE materials currently transported in UN1A2 Specification Package after the existing authorization expires, a replacement Type A(F) material packaging design is under development by the Savannah River National Laboratory. This paper presents a summary of the prototype design effort and testing of the new Type A(F) Package development for the DOE. This paper discusses the progress made in the development of a Type A Fissile Packaging to replace the expiring 49 CFR UN1A2 Specification Fissile Package. The Specification Package was mostly a single-use waste disposal container. The design requirements and authorized radioactive material contents of the UN1A2 Specification Package were defined in 49 CFR. A UN1A2 Specification Package was authorized to ship up to 350 grams of U-235 in any enrichment and in any non-pyrophoric form. The design was specified as a 55-gallon 1A2 drum overpack with a body constructed from 18 gauge steel with a 16 gauge drum lid. Drum closure was specified as a standard 12-gauge ring closure. The inner product container size was not specified but was listed as any container that met Specification 7A requirements per 49 CFR 178.350. Specification 7A containers were required to withstand Type A packaging tests required by 49CFR173.465 with compliance demonstrated through testing, analysis or similarity to other containers. The maximum weight of the 7A product container, the radioactive content, and any internal packaging was limited to 200 lbs. The total gross weight for the UN1A2 Specification Package was limited to 350 lbs. No additional restrictions were applied. Authorization for use did not require the UN1A2 Specification Package to be tested to the Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) required for performance based, Type A(F) packages certified by the NRC or DOE. The Type A(F) Packaging design discussed in this paper is required to be in compliance with the regulatory safety requirements defined in Code of Federal Regulations (CFR) 10 CFR 71.41 through 71.47 and 10 CFR71.71. Sub-criticality of content must be maintained under the Hypothetical Accident Conditions specified under 10 CFR71.73. These federal regulations, and other applicable DOE Orders and Guides, govern design requirements for a Type A(F) package. Type A(F) packages with less than an A2 quantity of radioactive material are not required to have a leak testable boundary. With this exception a Type A(F) package design is subject to the same test requirements set forth for the design of a performance based Type B packaging.

  4. Hanford Site radioactive hazardous materials packaging directory

    SciTech Connect (OSTI)

    McCarthy, T.L.

    1995-12-01

    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.

  5. Examples of Cost Estimation Packages

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

    1997-03-28

    Estimates can be performed in a variety of ways. Some of these are for projects for an undefined scope, a conventional construction project, or where there is a level of effort required to complete the work. Examples of cost estimation packages for these types of projects are described in this appendix.

  6. Safety evaluation for packaging (onsite) for the concrete-shielded RH TRU drum for the 327 Postirradiation Testing Laboratory

    SciTech Connect (OSTI)

    Smith, R.J.

    1998-03-31

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments. The drum will be used for transport of 327 Building legacy waste from the 300 Area to a solid waste storage facility on the Hanford Site.

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

    SciTech Connect (OSTI)

    Shappert, L.B.; Bowman, S.M.; Arnold, E.D.

    1998-08-01

    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.

  8. Waste Shipment Tracking Technology Lowers Costs, Increases Efficiency |

    Energy Savers [EERE]

    Department of Energy Waste Shipment Tracking Technology Lowers Costs, Increases Efficiency Waste Shipment Tracking Technology Lowers Costs, Increases Efficiency February 27, 2013 - 12:00pm Addthis This graphic shows how the radiofrequency identification technology tracks and monitors packages in transport, in-transit stops and storage. This graphic shows how the radiofrequency identification technology tracks and monitors packages in transport, in-transit stops and storage. The technology

  9. Radiolytic gas generation from cement-based waste hosts for DOE low-level radioactive wastes

    SciTech Connect (OSTI)

    Dole, L.R.; Friedman, H.A.

    1986-01-01

    Using cement-based immobilization binders with simulated radioactive waste containing sulfate, nitrate, nitrite, phosphate, and fluoride anions, the gamma- and alpha-radiolytic gas generation factors (G/sub t/, molecules/100 eV) and gas compositions were measured on specimens of cured grouts. These tests studied the effects of; (1) waste composition; (2) the sample surface-to-volume ratio; (3) the waste slurry particle size; and (4) the water content of the waste host formula. The radiolysis test vessels were designed to minimize the ''dead'' volume and to simulate the configuration of waste packages.

  10. Numerical recipes for mold filling simulation

    SciTech Connect (OSTI)

    Kothe, D.; Juric, D.; Lam, K.; Lally, B.

    1998-07-01

    Has the ability to simulate the filling of a mold progressed to a point where an appropriate numerical recipe achieves the desired results? If results are defined to be topological robustness, computational efficiency, quantitative accuracy, and predictability, all within a computational domain that faithfully represents complex three-dimensional foundry molds, then the answer unfortunately remains no. Significant interfacial flow algorithm developments have occurred over the last decade, however, that could bring this answer closer to maybe. These developments have been both evolutionary and revolutionary, will continue to transpire for the near future. Might they become useful numerical recipes for mold filling simulations? Quite possibly. Recent progress in algorithms for interface kinematics and dynamics, linear solution methods, computer science issues such as parallelization and object-oriented programming, high resolution Navier-Stokes (NS) solution methods, and unstructured mesh techniques, must all be pursued as possible paths toward higher fidelity mold filling simulations. A detailed exposition of these algorithmic developments is beyond the scope of this paper, hence the authors choose to focus here exclusively on algorithms for interface kinematics. These interface tracking algorithms are designed to model the movement of interfaces relative to a reference frame such as a fixed mesh. Current interface tracking algorithm choices are numerous, so is any one best suited for mold filling simulation? Although a clear winner is not (yet) apparent, pros and cons are given in the following brief, critical review. Highlighted are those outstanding interface tracking algorithm issues the authors feel can hamper the reliable modeling of today`s foundry mold filling processes.

  11. Departmental Materials Transportation and Packaging Management

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

    2010-11-18

    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.

  12. Amesos2 Templated Direct Sparse Solver Package

    Energy Science and Technology Software Center (OSTI)

    2011-05-24

    Amesos2 is a templated direct sparse solver package. Amesos2 provides interfaces to direct sparse solvers, rather than providing native solver capabilities. Amesos2 is a derivative work of the Trilinos package Amesos.

  13. TriUtils Trilinos Utilities Package

    Energy Science and Technology Software Center (OSTI)

    2011-09-26

    TriUtils is a package of utilities for other Trilinos packages. TriUtils contains utilities to perform common operations such as command line parsing, and input file reading.

  14. Safety Evaluation for Packaging for onsite Transfer of plutonium recycle test reactor ion exchange columns

    SciTech Connect (OSTI)

    Smith, R.J.

    1995-09-11

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the use of three U.S. Department of Transportation (DOT) 7A, Type A metal boxes (Capital Industries Part No. S 0600-0600-1080- 0104) to package 12 Plutonium Recycle Test Reactor (PRTR) ion exchange columns as low-level waste (LLW). The packages will be transferred from the 309 Building in the 300 Area to low level waste burial in the 200 West Area. Revision 1 of WHC-SD-TP-SEP-035 (per ECN No. 621467) documents that the boxes containing ion exchange columns and grout will maintain the payload under normal conditions of transport if transferred without the box lids

  15. Solution Package Scope Definition, Report 72, Salt Waste (SP #72)

    Broader source: Energy.gov [DOE]

    Supporting Technical Document for the Radiological Release Accident Investigation Report (Phase II Report)

  16. Cleanup Verification Package for the 116-K-2 Effluent Trench

    SciTech Connect (OSTI)

    J. M. Capron

    2006-04-04

    This cleanup verification package documents completion of remedial action for the 116-K-2 effluent trench, also referred to as the 116-K-2 mile-long trench and the 116-K-2 site. During its period of operation, the 116-K-2 site was used to dispose of cooling water effluent from the 105-KE and 105-KW Reactors by percolation into the soil. This site also received mixed liquid wastes from the 105-KW and 105-KE fuel storage basins, reactor floor drains, and miscellaneous decontamination activities.

  17. Temperature-package power correlations for open-mode geologic disposal concepts.

    SciTech Connect (OSTI)

    Hardin, Ernest L.

    2013-02-01

    Logistical simulation of spent nuclear fuel (SNF) management in the U.S. combines storage, transportation and disposal elements to evaluate schedule, cost and other resources needed for all major operations leading to final geologic disposal. Geologic repository reference options are associated with limits on waste package thermal power output at emplacement, in order to meet limits on peak temperature for certain key engineered and natural barriers. These package power limits are used in logistical simulation software such as CALVIN, as threshold requirements that must be met by means of decay storage or SNF blending in waste packages, before emplacement in a repository. Geologic repository reference options include enclosed modes developed for crystalline rock, clay or shale, and salt. In addition, a further need has been addressed for open modes in which SNF can be emplaced in a repository, then ventilated for decades or longer to remove heat, prior to permanent repository closure. For each open mode disposal concept there are specified durations for surface decay storage (prior to emplacement), repository ventilation, and repository closure operations. This study simulates those steps for several timing cases, and for SNF with three fuel-burnup characteristics, to develop package power limits at which waste packages can be emplaced without exceeding specified temperature limits many years later after permanent closure. The results are presented in the form of correlations that span a range of package power and peak postclosure temperature, for each open-mode disposal concept, and for each timing case. Given a particular temperature limit value, the corresponding package power limit for each case can be selected for use in CALVIN and similar tools.

  18. Microwave thawing package and method

    DOE Patents [OSTI]

    Fathi, Zakaryae; Lauf, Robert J.

    2004-03-16

    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.

  19. Notices 7004, Alternate Preservation, Packaging,

    Energy Savers [EERE]

    577 Federal Register / Vol. 79, No. 14 / Wednesday, January 22, 2014 / Notices 7004, Alternate Preservation, Packaging, and Packing and 252.211-7005, Substitutions for Military or Federal Specifications and Standards; OMB Control Number 0704-0398. Type of Request: Extension. Number of Respondents: 385. Responses Per Respondent: 1.4. Annual Responses: 573. Average Burden Per Response: Approximately 2 hours. Annual Burden Hours: 1,136. Needs and Uses: This information collection permits offers to-

  20. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA, JUNE 2006

    SciTech Connect (OSTI)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION NEVADA SITE OFFICE

    2006-06-01

    This document establishes the U.S. Department of Energy (DOE), 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 (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) 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 NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  1. Ultrafast carriers dynamics in filled-skutterudites

    SciTech Connect (OSTI)

    Guo, Liang; Xu, Xianfan; Salvador, James R.

    2015-06-08

    Carrier dynamics of filled-skutterudites, an important class of thermoelectric materials, is investigated using ultrafast optical spectroscopy. By tuning the wavelength of the probe laser, charge transfers at different electronic energy levels are interrogated. Analysis based on the Kramers-Kronig relation explains the complex spectroscopy data, which is mainly due to band filling caused by photo-excited carriers and free carrier absorption. The relaxation time of hot carriers is found to be about 0.4–0.6 ps, depending on the electronic energy level, and the characteristic time for carrier-phonon equilibrium is about 0.95 ps. These studies of carrier dynamics, which fundamentally determines the transport properties of thermoelectric material, can provide guidance for the design of materials.

  2. 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.

    1999-03-26

    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.

  3. RECLAMATION OF RADIOACTIVE MATERIAL PACKAGING COMPONENTS

    SciTech Connect (OSTI)

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

    2011-06-06

    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.

  4. Environmental evaluation of municipal waste prevention

    SciTech Connect (OSTI)

    Gentil, Emmanuel C.; Gallo, Daniele; Christensen, Thomas H.

    2011-12-15

    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.

  5. UWV (Unmanned Water Vehicle) - Umbra Package v. 1.0

    Energy Science and Technology Software Center (OSTI)

    2012-09-13

    This package contains modules that model the mobility of systems moving in the water. This package currently models first order physics -basically a velocity integrator. This package depends on interface classes (typically base classes) that reside in the Mobility package.

  6. EXAMINATION OF SHIPPING PACKAGE 9975-05050

    SciTech Connect (OSTI)

    Daugherty, W.

    2014-11-06

    Shipping package 9975-05050 was examined in K-Area following its identification as a high wattage package. Elevated temperature and fiberboard moisture content are key parameters that impact the degradation rate of fiberboard within 9975 packages in a storage environment. The high wattage of this package contributes significantly to component temperatures. After examination in K-Area, the package was provided to SRNL for further examination of the fiberboard assembly. The moisture content of the fiberboard was relatively low (compared to packages examined previously), but the moisture gradient (between fiberboard ID and OD surfaces) was relatively high, as would be expected for the high heat load. The cane fiberboard appeared intact and displayed no apparent change in integrity relative to a new package.

  7. Next-Generation LED Package Architectures Enabled by Thermally...

    Energy Savers [EERE]

    LED Package Architectures Enabled by Thermally Conductive Transparent Encapsulants Next-Generation LED Package Architectures Enabled by Thermally Conductive Transparent ...

  8. Solid Waste Processing Center Primary Opening Cells Systems, Equipment and Tools

    SciTech Connect (OSTI)

    Bailey, Sharon A.; Baker, Carl P.; Mullen, O Dennis; Valdez, Patrick LJ

    2006-04-17

    This document addresses the remote systems and design integration aspects of the development of the Solid Waste Processing Center (SWPC), a facility to remotely open, sort, size reduce, and repackage mixed low-level waste (MLLW) and transuranic (TRU)/TRU mixed waste that is either contact-handled (CH) waste in large containers or remote-handled (RH) waste in various-sized packages.

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

    SciTech Connect (OSTI)

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

    1991-10-11

    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.

  10. PORTABLE ACOUSTIC MONITORING PACKAGE (PAMP)

    SciTech Connect (OSTI)

    John l. Loth; Gary J. Morris; George M. Palmer; Richard Guiler; Deepak Mehra

    2003-07-01

    The 1st generation acoustic monitoring package was designed to detect and analyze weak acoustic signals inside natural gas transmission lines. Besides a microphone it housed a three-inch diameter aerodynamic acoustic signal amplifier to maximize sensitivity to leak induced {Delta}p type signals. The theory and test results of this aerodynamic signal amplifier was described in the master's degree thesis of our Research Assistant Deepak Mehra who is about to graduate. To house such a large three-inch diameter sensor required the use of a steel 300-psi rated 4 inch weld neck flange, which itself weighed already 29 pounds. The completed 1st generation Acoustic Monitoring Package weighed almost 100 pounds. This was too cumbersome to mount in the field, on an access port at a pipeline shut-off valve. Therefore a 2nd generation and truly Portable Acoustic Monitor was built. It incorporated a fully self-contained {Delta}p type signal sensor, rated for line pressures up to 1000 psi with a base weight of only 6 pounds. This is the Rosemont Inc. Model 3051CD-Range 0, software driven sensor, which is believed to have industries best total performance. Its most sensitive unit was purchased with a {Delta}p range from 0 to 3 inch water. This resulted in the herein described 2nd generation: Portable Acoustic Monitoring Package (PAMP) for pipelines up to 1000 psi. Its 32-pound total weight includes an 18-volt battery. Together with a 3 pound laptop with its 4-channel data acquisition card, completes the equipment needed for field acoustic monitoring of natural gas transmission pipelines.

  11. Trends in packaged steam generators

    SciTech Connect (OSTI)

    Ganapathy, V. [ABCO Industries, Abilene, TX (United States)

    1996-09-01

    Oil and gas-fired packaged steam generators are used in many industrial plants. They generate saturated or superheated steam up to 250,000 lb/hr, 1000 psig, and 950 F. They may be used for continuous steam generation or as standby boilers in cogeneration systems. Numerous variables affect the design of this equipment. A few important considerations should be addressed at an early point by the plant engineer specifying or evaluating equipment options. These considerations include trends such as customized designs that minimize operating costs and ensure emissions regulations are met. The paper discusses efficiency considerations first.

  12. Block LancZos PACKage

    Energy Science and Technology Software Center (OSTI)

    2005-05-01

    BLZPACK (for Block LancZos PACKage) is a standard Fortran 77 implementation of the block Lanczos algorithm intended for the solution of the standard eigenvalue problem Ax=ux or the generalized eigenvalue problem Ax=uBx, where A and B are real, sparse symmetric matrices, u and eigenvalue and x and eigenvector. The development of this eigensolver was motivated by the need to solve large, sparse, generalized problems from free vibration analyses in structural engineering. Several upgrades were performedmoreĀ Ā» afterwards aiming at the solution of eigenvalues problems from a wider range of applications.Ā«Ā less

  13. 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.

    1999-09-24

    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.

  14. Liquid-Filled Piping System Analysis

    Energy Science and Technology Software Center (OSTI)

    1993-07-07

    WHAM6 is used to calculate pressure and velocity transients in liquid-filled piping networks. It can be applied to multiloop complex piping networks consisting of dead ends, elbows, orifices, multiple-branch tees, changes of flow passage cross section, check valves, pumps, pressurizers or tanks, and exit valves or breaks. Hydraulic losses are considered. Transients can be initiated either by closure or opening of one or more exit valves (equivalent to system ruptures) or by a prescribed gasmoreĀ Ā» pressure history in a pressurizer tank.Ā«Ā less

  15. Waste Hoist

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

    Disposal Waste Disposal Trucks transport debris from Oak RidgeĆ¢Ā€Ā™s cleanup sites to the onsite CERCLA disposal area, the Environmental Management Waste Management Facility. Trucks transport debris from Oak Ridge's cleanup sites to the onsite CERCLA disposal area, the Environmental Management Waste Management Facility. The low-level radiological and hazardous wastes generated from Oak Ridge's cleanup projects are disposed in the Environmental Management Waste Management Facility (EMWMF). The

  16. Summary of Uranium Solubility Studies in Concrete Waste Forms and Vadose Zone Environments

    SciTech Connect (OSTI)

    Golovich, Elizabeth C.; Wellman, Dawn M.; Serne, R. Jeffrey; Bovaird, Chase C.

    2011-09-30

    One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and 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. This report presents the results of investigations elucidating the uranium mineral phases controlling the long-term fate of uranium within concrete waste forms and the solubility of these phases in concrete pore waters and alkaline, circum-neutral vadose zone environments.

  17. Waste reduction through consumer education. Final report

    SciTech Connect (OSTI)

    Harrison, E.Z.

    1996-05-01

    The Waste Reduction through Consumer Education research project was conducted to determine how environmental educational strategies influence purchasing behavior in the supermarket. The objectives were to develop, demonstrate, and evaluate consumer education strategies for waste reduction. The amount of waste generated by packaging size and form, with an adjustment for local recyclability of waste, was determined for 14 product categories identified as having more waste generating and less waste generating product choices (a total of 484 products). Using supermarket scan data and shopper identification numbers, the research tracked the purchases of shoppers in groups receiving different education treatments for 9 months. Statistical tests applied to the purchase data assessed patterns of change between the groups by treatment period. Analysis of the data revealed few meaningful statistical differences between study groups or changes in behavior over time. Findings suggest that broad brush consumer education about waste reduction is not effective in changing purchasing behaviors in the short term. However, it may help create a general awareness of the issues surrounding excess packaging and consumer responsibility. The study concludes that the answer to waste reduction in the future may be a combination of voluntary initiatives by manufacturers and retailers, governmental intervention, and better-informed consumers.

  18. Packaging of electro-microfluidic devices

    DOE Patents [OSTI]

    Benavides, Gilbert L.; Galambos, Paul C.; Emerson, John A.; Peterson, Kenneth A.; Giunta, Rachel K.; Watson, Robert D.

    2002-01-01

    A new architecture for packaging surface micromachined electro-microfluidic devices is presented. This architecture relies on two scales of packaging to bring fluid to the device scale (picoliters) from the macro-scale (microliters). The architecture emulates and utilizes electronics packaging technology. The larger package consists of a circuit board with embedded fluidic channels and standard fluidic connectors (e.g. Fluidic Printed Wiring Board). The embedded channels connect to the smaller package, an Electro-Microfluidic Dual-Inline-Package (EMDIP) that takes fluid to the microfluidic integrated circuit (MIC). The fluidic connection is made to the back of the MIC through Bosch-etched holes that take fluid to surface micromachined channels on the front of the MIC. Electrical connection is made to bond pads on the front of the MIC.

  19. Packaging of electro-microfluidic devices

    DOE Patents [OSTI]

    Benavides, Gilbert L.; Galambos, Paul C.; Emerson, John A.; Peterson, Kenneth A.; Giunta, Rachel K.; Zamora, David Lee; Watson, Robert D.

    2003-04-15

    A new architecture for packaging surface micromachined electro-microfluidic devices is presented. This architecture relies on two scales of packaging to bring fluid to the device scale (picoliters) from the macro-scale (microliters). The architecture emulates and utilizes electronics packaging technology. The larger package consists of a circuit board with embedded fluidic channels and standard fluidic connectors (e.g. Fluidic Printed Wiring Board). The embedded channels connect to the smaller package, an Electro-Microfluidic Dual-Inline-Package (EMDIP) that takes fluid to the microfluidic integrated circuit (MIC). The fluidic connection is made to the back of the MIC through Bosch-etched holes that take fluid to surface micromachined channels on the front of the MIC. Electrical connection is made to bond pads on the front of the MIC.

  20. Lessons learned during Type A Packaging testing

    SciTech Connect (OSTI)

    O`Brien, J.H.; Kelly, D.L.

    1995-11-01

    For the past 6 years, the US Department of Energy (DOE) Office of Facility Safety Analysis (EH-32) has contracted Westinghouse Hanford Company (WHC) to conduct compliance testing on DOE Type A packagings. The packagings are tested for compliance with the U.S. Department of Transportation (DOT) Specification 7A, general packaging, Type A requirements. The DOE has shared the Type A packaging information throughout the nuclear materials transportation community. During testing, there have been recurring areas of packaging design that resulted in testing delays and/or initial failure. The lessons learned during the testing are considered a valuable resource. DOE requested that WHC share this resource. By sharing what is and can be encountered during packaging testing, individuals will hopefully avoid past mistakes.

  1. Thermoelectric Properties of Indium-Filled Skutterudites

    SciTech Connect (OSTI)

    He, Tao; Chen, Jiazhong; Rosenfeld, H. David; Subramanian, M.A.

    2008-09-18

    Structural, electrical, and thermal transport properties of CoSb{sub 3} partially filled with indium are reported. Polycrystalline samples of In{sub x}Co{sub 4}Sb{sub 12} (0 {le} x {le} 0.3) were prepared by solid-state reaction under a gas mixture of 5% H{sub 2} and 95% Ar. The solubility limit of the indium filling voids in CoSb{sub 3} was found to be close to 0.22. Synchrotron X-ray diffraction refinement of the x = 0.2 sample showed that the indium is located in the classic rattler site and has a substantially larger thermal factor than those of Co and Sb. The electrical resistivity, Seebeck coefficients, and thermal conductivity of the In{sub x}Co{sub 4}Sb{sub 12} samples were measured in the temperature range of 300-600 K. All samples showed metal-like behavior, and the large negative Seebeck coefficients indicated n-type conduction. The thermal conductivity decreased with increasing temperature for all samples. A thermoelectric figure-of-merit (ZT) {ge} 1 (n-type) has been achieved when x {ge} 0.2 in In{sub x}Co{sub 4}Sb{sub 12} at 575 K.

  2. NFR TRIGA package design review report

    SciTech Connect (OSTI)

    Clements, M.D.

    1994-08-26

    The purpose of this document is to compile, present and document the formal design review of the NRF TRIGA packaging. The contents of this document include: the briefing meeting presentations, package description, design calculations, package review drawings, meeting minutes, action item lists, review comment records, final resolutions, and released drawings. This design review required more than two meeting to resolve comments. Therefore, there are three meeting minutes and two action item lists.

  3. Single Packaged Vertical Units | Department of Energy

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

    Single Packaged Vertical Units Single Packaged Vertical Units The Department of Energy (DOE) develops standardized data templates for reporting the results of tests conducted in accordance with current DOE test procedures. Templates may be used by third-party laboratories under contract with DOE that conduct testing in support of ENERGY STARĀ® verification, DOE rulemakings, and enforcement of the federal energy conservation standards. File Single Packaged Vertical Units -- v2.0 More Documents

  4. Cost Comparison for the Transfer of Select Calcined Waste Canisters to the Monitored Geologic Repository at Yucca Mountain, NV

    SciTech Connect (OSTI)

    Michael B. Heiser; Clark B. Millet

    2005-10-01

    This report performs a life-cycle cost comparison of three proposed canister designs for the shipment and disposition of Idaho National Laboratory high-level calcined waste currently in storage at the Idaho Nuclear Technology and Engineering Center to the proposed national monitored geologic repository at Yucca Mountain, Nevada. Concept A (2 × 10-ft) and Concept B (2 × 15-ft) canisters are comparable in design, but they differ in size and waste loading options and vary proportionally in weight. The Concept C (5.5 × 17.5-ft) canister (also called the “super canister”), while similar in design to the other canisters, is considerably larger and heavier than Concept A and B canisters and has a greater wall thickness. This report includes estimating the unique life-cycle costs for the three canister designs. Unique life-cycle costs include elements such as canister purchase and filling at the Idaho Nuclear Technology and Engineering Center, cask preparation and roundtrip consignment costs, final disposition in the monitored geologic repository (including canister off-loading and placement in the final waste disposal package for disposition), and cask purchase. Packaging of the calcine "as-is" would save $2.9 to $3.9 billion over direct vitrification disposal in the proposed national monitored geologic repository at Yucca Mountain, Nevada. Using the larger Concept C canisters would use 0.75 mi less of tunnel space, cost $1.3 billion less than 10-ft canisters of Concept A, and would be complete in 6.2 years.

  5. Getting waste ready for shipment to the WIPP: integration of characterization and certification activities

    SciTech Connect (OSTI)

    Sinkule, B.; Knudsen, K.; Rogers, P.

    1996-06-01

    The Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC) serve as the primary directive for assuring the safe handling, transportation, and disposal of transuranic (TRU) waste generated at Department of Energy (DOE) sites. The WIPP WAC address fulfillment of WIPP`s operational safety and performance assessment criteria, compliance with Resource Conservation and Recovery Act (RCRA) requirements, and preparation of waste packages that meet all transportation criteria. At individual generator sites, preparation of transuranic waste for final disposal at WIPP includes characterizing the waste to meet the requirements of the transuranic Waste Characterization Quality Assurance Program Plan (QAPP) and certifying waste containers to meet the WIPP WAC and the Transuranic Package Transporter-II Authorized Methods for Payload Control (TRAMPAC). This paper compares the quality assurance and quality control requirements specified in the WIPP WAC, QAPP, and TRAMPAC and discusses the potential to consolidate activities to comply with the TRU waste characterization and certification program requirements.

  6. DOE Identifies its Preferred Alternative for Certain Hanford Tank Wastes

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) is announcing its preferred alternative for wastes contained in underground radioactive waste storage tanks evaluated in the Final Tank Closure and Waste Management Environmental Impact Statement for the Hanford Site, Richland, Washington (Final TC & WM EIS, DOE/EIS-0391, December 2012). With regard to those wastes that, in the future, may be properly and legally classified as mixed transuranic waste (mixed TRU waste). DOE's preferred alternative is to retrieve, treat, package, and characterize and certify the wastes for disposal at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico, a geologic repository for the disposal of mixed TRU waste generated by atomic energy defense activities.

  7. Single Packaged Vertical Units | Department of Energy

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

    standards. File Single Packaged Vertical Units -- v2.0 More Documents & Publications Room Air Conditioners Commercial Refrigeration Equipment Commercial Refrigeration Equipment

  8. Combustion Analysis Software Package for Internal Combustion...

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

    Return to Search Combustion Analysis Software Package for Internal Combustion Engines Colorado State University Contact CSU About This Technology Technology Marketing Summary ...

  9. Microelectronic device package with an integral window

    DOE Patents [OSTI]

    Peterson, Kenneth A.; Watson, Robert D.

    2002-01-01

    An apparatus for packaging of microelectronic devices, including an integral window. The microelectronic device can be a semiconductor chip, a CCD chip, a CMOS chip, a VCSEL chip, a laser diode, a MEMS device, or a IMEMS device. The package can include a cofired ceramic frame or body. The package can have an internal stepped structure made of one or more plates, with apertures, which are patterned with metallized conductive circuit traces. The microelectronic device can be flip-chip bonded on the plate to these traces, and oriented so that the light-sensitive side is optically accessible through the window. A cover lid can be attached to the opposite side of the package. The result is a compact, low-profile package, having an integral window that can be hermetically-sealed. The package body can be formed by low-temperature cofired ceramic (LTCC) or high-temperature cofired ceramic (HTCC) multilayer processes with the window being simultaneously joined (e.g. cofired) to the package body during LTCC or HTCC processing. Multiple chips can be located within a single package. The cover lid can include a window. The apparatus is particularly suited for packaging of MEMS devices, since the number of handling steps is greatly reduced, thereby reducing the potential for contamination.

  10. Hazardous Material Packaging for Transport - Administrative Procedures

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

    1986-09-30

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

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

    SciTech Connect (OSTI)

    1994-10-01

    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.

  12. HMPT: Hazardous Waste Transportation Live 27928, Test 27929 (Technical

    Office of Scientific and Technical Information (OSTI)

    Report) | SciTech Connect HMPT: Hazardous Waste Transportation Live 27928, Test 27929 Citation Details In-Document Search Title: HMPT: Hazardous Waste Transportation Live 27928, Test 27929 HMPT: Hazardous Waste Transportation (Live 27928, suggested one time and associated Test 27929, required initially and every 36 months) addresses the Department of Transportation (DOT) function-specific training requirements of the hazardous materials packagings and transportation (HMPT) Los Alamos

  13. On-site waste storage assuring the success of on-site, low-level nuclear waste storage

    SciTech Connect (OSTI)

    Preston, E.L.

    1986-09-21

    Waste management has reached paramount importance in recent years. The successful management of radioactive waste is a key ingredient in the successful operation of any nuclear facility. This paper discusses the options available for on-site storage of low-level radioactive waste and those options that have been selected by the Department of Energy facilities operated by Martin Marietta Energy Systems, Inc. in Oak Ridge, Tennessee. The focus of the paper is on quality assurance (QA) features of waste management activities such as accountability and retrievability of waste materials and waste packages, retrievability of data, waste containment, safety and environmental monitoring. Technical performance and careful documentation of that performance are goals which can be achieved only through the cooperation of numerous individuals from waste generating and waste managing organizations, engineering, QA, and environmental management.

  14. PRIDE Surveillance Projects Data Packaging Project Information Package Specification Version 1.1

    SciTech Connect (OSTI)

    Kelleher, D. M.; Shipp, R. L.; Mason, J. D.

    2010-08-31

    Information Package Specification version 1.1 describes an XML document format called an information package that can be used to store information in information management systems and other information archives. An information package consists of package information, the context required to understand and use that information, package metadata that describes the information, and XML signatures that protect the information. The information package described in this specification was designed to store Department of Energy (DOE) and National Nuclear Security Administration (NNSA) information and includes the metadata required for that information: a unique package identifier, information marking that conforms to DOE and NNSA requirements, and access control metadata. It is an implementation of the Open Archival Information System (OAIS) Reference Model archival information package tailored to meet NNSA information storage requirements and designed to be used in the computing environments at the Y-12 National Security Complex and at other NNSA sites.

  15. Reference commercial high-level waste glass and canister definition.

    SciTech Connect (OSTI)

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

    1981-09-01

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

  16. Belos Block Linear Solvers Package

    Energy Science and Technology Software Center (OSTI)

    2004-03-01

    Belos is an extensible and interoperable framework for large-scale, iterative methods for solving systems of linear equations with multiple right-hand sides. The motivation for this framework is to provide a generic interface to a collection of algorithms for solving large-scale linear systems. Belos is interoperable because both the matrix and vectors are considered to be opaque objects--only knowledge of the matrix and vectors via elementary operations is necessary. An implementation of Balos is accomplished viamoreĀ Ā» the use of interfaces. One of the goals of Belos is to allow the user flexibility in specifying the data representation for the matrix and vectors and so leverage any existing software investment. The algorithms that will be included in package are Krylov-based linear solvers, like Block GMRES (Generalized Minimal RESidual) and Block CG (Conjugate-Gradient).Ā«Ā less

  17. Gas Generation Rates as an Indicator for the Long Term Stability of Radioactive Waste Products

    SciTech Connect (OSTI)

    Steyer, S.; Brennecke, P.; Bandt, G.; Kroger, H.

    2007-07-01

    Pursuant to the 'Act on the Peaceful Utilization of Atomic Energy and the Protection against its Hazards' (Atomic Energy Act) the Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz, BfS) is legally responsible for the construction and operation of federal facilities for the disposal of radioactive waste. Within the scope of this responsibility, particular due to par. 74(1) Ordinance on Radiation Protection, BfS defines all safety-related requirements on waste packages envisaged for disposal, establishes guidelines for the conditioning of radioactive waste and approves the fulfillment of the waste acceptance requirements within the radioactive waste quality control system. BfS also provides criteria to enable the assessment of methods for the treatment and packaging of radioactive waste to produce waste packages suitable for disposal according to par. 74(2) Ordinance on Radiation Protection. Due to the present non-availability of a repository in Germany, quality control measures for all types of radioactive waste products are carried out prior to interim storage with respect to the future disposal. As a result BfS approves the demonstrated properties of the radioactive waste packages and confirms the fulfillment of the respective requirements. After several years of storage the properties of waste packages might have changed. By proving, that such changes have no significant impact on the quality of the waste product, the effort of requalification could be minimized. Therefore, data on the long-term behavior of radioactive waste products need to be acquired and indicators to prove the long-term stability have to be quantified. Preferably, such indicators can be determined easily with non-destructive methods, even for legacy waste packages. A promising parameter is the gas generation rate. The relationship between gas generation rate and long term stability is presented as first result of an ongoing study on behalf of BfS. Permissible gas generation rates that ensure adequate product stability with respect to future disposal are to be identified. (authors)

  18. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    G. Radulesscu; J.S. Tang

    2000-06-07

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

  19. Application of EPA regulations to low-level radioactive waste

    SciTech Connect (OSTI)

    Bowerman, B.S.; Piciulo, P.L.

    1985-01-01

    The survey reported here was conducted with the intent of identifying categories of low-level radioactive wastes which would be classified under EPA regulations 40 CFR Part 261 as hazardous due to the chemical properties of the waste. Three waste types are identified under these criteria as potential radioactive mixed wastes: wastes containing organic liquids; wastes containing lead metal; and wastes containing chromium. The survey also indicated that certain wastes, specific to particular generators, may also be radioactive mixed wastes. Ultimately, the responsibility for determining whether a facility's wastes are mixed wastes rest with the generator. However, the uncertainties as to which regulations are applicable, and the fact that no legal definition of mixed wastes exists, make such a determination difficult. In addition to identifying mixed wastes, appropriate methods for the management of mixed wastes must be defined. In an ongoing study, BNL is evaluating options for the management of mixed wastes. These options will include segregation, substitution, and treatments to reduce or eliminate chemical hazards associated with the wastes listed above. The impacts of the EPA regulations governing hazardous wastes on radioactive mixed waste cannot be assessed in detail until the applicability of these regulations is agreed upon. This issue is still being discussed by EPA and NRC and should be resolved in the near future. Areas of waste management which may affect generators of mixed wastes include: monitoring/tracking of wastes before shipment; chemical testing of wastes; permits for treatment of storage of wastes; and additional packaging requirements. 3 refs., 1 fig., 2 tabs.

  20. Aging Studies of Filled and Unfilled VCE

    SciTech Connect (OSTI)

    Letant, S; Herberg, J; Alviso, C; Small, W; Mulcahy, H; Pearson, M; Wilson, T; Chinn, S; Maxwell, R

    2009-11-10

    This report presents data on the effects of temperature and gamma radiation on the chemical and structural properties of both filled and unfilled VCE material produced by the Kansas City Plant using WR-qualified processes. Thermal effects up to 300 C and gamma irradiation doses of 1 MRad and 25 MRad were investigated under atmospheric conditions. Characterization techniques used in the study comprise Thermogravimetric Analysis (TGA), Differential Scanning Calorimetry (DSC), X-ray Diffraction (XRD), Tensile Testing, Solid Phase MicroExtraction - Gas Chromatography/Mass Spectrometry (SPME-GC/MS), phenol extraction followed by HPLC, and various Nuclear Magnetic Resonance (NMR) techniques including: {sup 13}C, {sup 13}C {l_brace}{sup 1}H{r_brace} cross polarization (CP), {sup 1}H magic angle spinning (MAS), 13C{l_brace}{sup 1}H{r_brace} Wide-line-Separation (2D-WISE) and development of Center band-Only Detection of Exchange (CODEX).

  1. Constrained ceramic-filled polymer armor

    DOE Patents [OSTI]

    Sandstrom, D.J.; Calkins, N.C.; Gac, F.D.

    1990-11-13

    An armor system is disclosed in which a plurality of constraint cells are mounted on a surface of a substrate, which is metal armor plate or a similar tough material, such that the cells almost completely cover the surface of the substrate. Each constraint cell has a projectile-receiving wall parallel to the substrate surface and has sides which are perpendicular to and surround the perimeter of the receiving wall. The cells are mounted such that, in one embodiment, the substrate surface serves as a sixth side or closure for each cell. Each cell has inside of it a plate, termed the front plate, which is parallel to and in contact with substantially all of the inside surface of the receiving wall. The balance of each cell is completely filled with a projectile-abrading material, which is a ceramic material in particulate form dispersed in a polymeric matrix. 5 figs.

  2. Constrained ceramic-filled polymer armor

    DOE Patents [OSTI]

    Sandstrom, Donald J.; Calkins, Noel C.; Gac, Frank D.

    1990-01-01

    An armor system in which a plurality of constraint cells are mounted on a surface of a substrate, which is metal armor plate or a similar tough material, such that the cells almost completely cover the surface of the substrate. Each constraint cell has a projectile-receiving wall parallel to the substrate surface and has sides which are perpendicular to and surround the perimeter of the receiving wall. The cells are mounted such that, in one embodiment, the substrate surface serves as a sixth side or closure for each cell. Each cell has inside of it a plate, termed the front plate, which is parallel to and in contact with substantially all of the inside surface of the receiving wall. The balance of each cell is completely filled with a projectile-abrading material, which is a ceramic material in particulate form dispersed in a polymeric matrix.

  3. Method of filling a microchannel separation column

    DOE Patents [OSTI]

    Arnold, Don W. (Livermore, CA)

    2002-01-01

    A method for packing a stationary phase into a small diameter fluid passageway or flow channel. Capillary action is employed to distribute a stationary phase uniformly along both the length and diameter of the flow channel. The method disclosed here: 1) eliminates the need for high pressure pumps and fittings and the safety hazards associated therewith; 2) allows the use of readily available commercial microparticles, either coated or uncoated, as the stationary phase; 3) provides for different types of particles, different particle sizes, and different particle size distributions to be packed in sequence, or simultaneously; 4) eliminates the need for plugging the flow channel prior to adding the stationary phase to retain the packing particles; and 5) many capillaries can be filled simultaneously.

  4. K Basin sludge packaging design criteria (PDC) and safety analysis report for packaging (SARP) approval plan

    SciTech Connect (OSTI)

    Brisbin, S.A.

    1996-03-06

    This document delineates the plan for preparation, review, and approval of the Packaging Design Crieteria for the K Basin Sludge Transportation System and the Associated on-site Safety Analysis Report for Packaging. The transportation system addressed in the subject documents will be used to transport sludge from the K Basins using bulk packaging.

  5. TANK 18 AND 19-F TIER 1A EQUIPMENT FILL MOCK UP TEST SUMMARY

    SciTech Connect (OSTI)

    Stefanko, D.; Langton, C.

    2011-11-04

    The United States Department of Energy (US DOE) has determined that Tanks 18-F and 19-F have met the F-Tank Farm (FTF) General Closure Plan Requirements and are ready to be permanently closed. The high-level waste (HLW) tanks have been isolated from FTF facilities. To complete operational closure they will be filled with grout for the purpose of: (1) physically stabilizing the tanks, (2) limiting/eliminating vertical pathways to residual waste, (3) discouraging future intrusion, and (4) providing an alkaline, chemical reducing environment within the closure boundary to control speciation and solubility of select radionuclides. Bulk waste removal and heel removal equipment remain in Tanks 18-F and 19-F. This equipment includes the Advance Design Mixer Pump (ADMP), transfer pumps, transfer jets, standard slurry mixer pumps, equipment-support masts, sampling masts, dip tube assemblies and robotic crawlers. The present Tank 18 and 19-F closure strategy is to grout the equipment in place and eliminate vertical pathways by filling voids in the equipment to vertical fast pathways and water infiltration. The mock-up tests described in this report were intended to address placement issues identified for grouting the equipment that will be left in Tank 18-F and Tank 19-F. The Tank 18-F and 19-F closure strategy document states that one of the Performance Assessment (PA) requirements for a closed tank is that equipment remaining in the tank be filled to the extent practical and that vertical flow paths 1 inch and larger be grouted. The specific objectives of the Tier 1A equipment grout mock-up testing include: (1) Identifying the most limiting equipment configurations with respect to internal void space filling; (2) Specifying and constructing initial test geometries and forms that represent scaled boundary conditions; (3) Identifying a target grout rheology for evaluation in the scaled mock-up configurations; (4) Scaling-up production of a grout mix with the target rheology (16 second flow cone value) from 0.25 cubic feet to 4.3 cubic feet. (Ten 0.43 cubic batches were produced because full-scale equipment was not available for the Tier 1A test.); (5) Demonstrating continuous gravity filling of the ADMP mock up test form; (6) Demonstrating continuous gravity filling of 1 inch and 2 inch schedule 40 pipe; and (7) Demonstrating filling of 1 inch and 2 inch schedule 40 pipe from the bottom up by discharging through a tube inserted into the pipes. The Tier 1A mock-up test focused on the ADMP and pipes at least one inch in diameter. The ADMP which is located in center riser of Tank 18-F is a concern because the column for this long-shaft (55 ft) pump is unique and modification to the pump prior to placing it in service limited the flow path options for filling by creating a single flow path for filling and venting the ADMP support column. The large size, vertical orientation, and complicated flow path in the ADMP warrants a detailed description of this piece of ancillary equipment.

  6. KLU2 Direct Linear Solver Package

    Energy Science and Technology Software Center (OSTI)

    2012-01-04

    KLU2 is a direct sparse solver for solving unsymmetric linear systems. It is related to the existing KLU solver, (in Amesos package and also as a stand-alone package from University of Florida) but provides template support for scalar and ordinal types. It uses a left looking LU factorization method.

  7. PyTrilinos Rapid Prototyping Package

    Energy Science and Technology Software Center (OSTI)

    2005-03-01

    PyTrilinos provides access to selected Trilinos packages from the python scripting language. This allows interactive and dynamic creation of Trilinos objects, rapid prototyping that does not require compilation, and "gluing" Trilinos scripts to other python modules, such as plotting, etc. The currently supported packages are Epetra, EpetraExt, and NOX.

  8. Departmental Materials Transportation and Packaging Management

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

    2004-12-22

    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

  9. Nevada Test Site Waste Acceptance Criteria (NTSWAC), Rev. 7-01

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2009-05-01

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Nevada Test Site Waste Acceptance Criteria (NTSWAC). The NTSWAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NTSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex for disposal.

  10. A model waste analysis plan for commercial BIF facilities

    SciTech Connect (OSTI)

    Gossmann, D.; Woodford, J.

    1997-12-31

    EPA`s guidance, to date, of Waste Analysis Plans has failed to provide specific recommendations for the special needs of commercial BIF facilities. EPA has instead focused primarily on traditional incinerator and landfill options for hazardous waste management. In order to fill this gap, a model waste analysis plan for commercial BIF facilities is developed along with the rationale for key features. 5 tabs.

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

    DOE Patents [OSTI]

    Cowan, Richard G.; Blasewitz, Albert G.

    1982-01-01

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

  12. Improved thermoelectric performance of n-type Ca and Ca-Ce filled skutterudites

    SciTech Connect (OSTI)

    Thompson, Daniel R.; Liu, Chang; Ellison, Nicole D.; Salvador, James R.; Meyer, Martin S.; Haddad, Daad B.; Wang, Hsin; Cai, W.

    2014-12-28

    Thermoelectric (TE) technology for use in automotive waste heat recovery is being advanced by General Motors with support from the US Department of Energy. Skutterudites are a very promising material for this application of TE technology due to their superior mechanical properties and good TE performance. Double-filled Yb{sub x}Ba{sub y}Co{sub 4}Sb{sub 12} with ZT values around 1.1 at 750?K are the best performing n-type skutterudites produced on a large scale using an economically viable approach of melt spinning (MS) in conjunction with spark plasma sintering (SPS). Another economical production method on the tons scale, the melt quench annealing (MQA) technique, has been recently claimed by Treibacher Industrie AG, further information is available [G. Rogl et al., Acta Mater. 76, 434–448 (2014)]. A possible hurdle to commercial implementation of these materials is the use of rare earths as the fillers to reduce thermal conductivity and improve the electrical transport properties. It will be shown herein that skutterudites double-filled with Ca and Ce, both of which are lower-cost fillers, display markedly different TE properties depending on whether they are produced by MQA or MS + SPS synthesis techniques. Ca and Ce double-filled skutterudites prepared by MS + SPS have TE properties that are superior to the same compositions prepared by MQA and that are comparable to the best performing Yb and Ba filled materials. Furthermore, the results of this study suggest that the unusually poor transport properties of MQA Ca-filled skutterudites can be ascribed to deleterious secondary phases, which is contrary to reports in the literature attempting to explain these irregularities via band structure features.

  13. TANKS 18 AND 19-F EQUIPMENT GROUT FILL MATERIAL EVALUATION AND RECOMMENDATIONS

    SciTech Connect (OSTI)

    Stefanko, D.; Langton, C.

    2011-12-15

    The United States Department of Energy (US DOE) intends to remove Tanks 18-F and 19-F at the Savannah River Site (SRS) from service. The high-level waste (HLW) tanks have been isolated from the F-area Tank Farm (FTF) facilities and will be filled with cementitious grout for the purpose of: (1) physically stabilizing the empty volumes in the tanks, (2) limiting/eliminating vertical pathways from the surface to residual waste on the bottom of the tanks, (3) providing an intruder barrier, and (4) providing an alkaline, chemical reducing environment within the closure boundary to limit solubility of residual radionuclides. Bulk waste and heel waste removal equipment will remain in Tanks 18-F and 19-F when the tanks are closed. This equipment includes: mixer pumps, transfer pumps, transfer jets, equipment support masts, sampling masts and dip tube assemblies. The current Tank 18-F and 19-F closure strategy is to grout the internal void spaces in this equipment to eliminate fast vertical pathways and slow water infiltration to the residual material on the tank floor. This report documents the results of laboratory testing performed to identify a grout formulation for filling the abandoned equipment in Tanks 18-F and 19-F. The objective of this work was to formulate a flowable grout for filling internal voids of equipment that will remain in Tanks 18-F and 19-F during the final closures. This work was requested by V. A. Chander, Tank Farm Closure Engineering, in HLW-TTR-2011-008. The scope for this task is provided in the Task Technical and Quality Assurance Plan (TTQAP), SRNL-RP-2011-00587. The specific objectives of this task were to: (1) Prepare and evaluate the SRR cooling coil grout identified in WSRC-STI-2008-00298 per the TTR for this work. The cooling coil grout is a mixture of BASF MasterFlow{reg_sign} 816 cable grout (67.67 wt. %), Grade 100 ground granulated blast furnace slag (7.52 wt. %) and water (24.81 wt. %); (2) Identify equipment grout placement and performance properties; (3) Design up to 2 additional grout systems for filling the Tank 18-F and Tank 19-F equipment; (4) Prepare samples of candidate grouts and measure fresh properties, thermal properties and cured properties; (5) Recommend a grout for the Tier 1A equipment fill mock up - ADMP 4 foot high mock up, 1 inch and 2 inch pipes; (6) Support procurement of materials for the Tier 1A equipment fill mock up test; (7) Prepare samples of the recommended grout for hydraulic property measurements which can be used for comparison to values used in the F- Tank Farm Performance Assessment (PA); and (8) Document equipment fill grout data and recommendations in a report.

  14. DOE Announces Preference for Disposal of Hanford Transuranic Tank Waste at WIPP

    Broader source: Energy.gov [DOE]

    WASHINGTON, D.C. ā€“Ā  Today the U.S. Department of Energy (DOE) announced its preferred alternative to retrieve, treat, package, characterize and certify certain Hanford tank waste for disposal at...

  15. Overview of FordĀs Thermoelectric Programs: Waste Heat Recovery and Climate Control

    Broader source: Energy.gov [DOE]

    Overview of progress in TE waste heat recovery from sedan gasoline-engine exhaust, TE HVAC system in hybrid sedan, and establishing targets for cost, power density, packaging, durability, and systems integration

  16. Expected brine movement at potential nuclear waste repository salt sites

    SciTech Connect (OSTI)

    McCauley, V.S.; Raines, G.E.

    1987-08-01

    The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m/sup 3/ brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs.

  17. Handbook for Gas Filled RF Cavity Aficionados'

    SciTech Connect (OSTI)

    Tollestrup, A.V.; Chung, Moses; Yonehara, Katsuya; /Fermilab

    2009-05-01

    The use of hydrogen gas filled RF cavities in muon cooling channels has been proposed by Rolland Johnson. Impressive results have been obtained toward attaining high voltage gradients and rapid training in preliminary tests done at the FNAL MTA facility. However, so far it has not been possible to test them under conditions where they were subject to the transversal of a high intensity particle beam. This note is an attempt to bring together a description of some of the pertinent physical processes that take place in the dilute plasma that is generated in the hydrogen gas by the beam. Two effects dominate. The first is that the free electrons generated can load down the cavity and transfer its energy to heating the gas. The second is a question of what happens to the plasma in the longer term. There is an enormous literature on the subject of the subject of dilute hydrogen plasmas and we can tap into this information in order to understand and predict the behavior of the cavity.

  18. Safety evaluation for packaging (onsite) for the Pacific Northwest National Laboratory HEPA filter box

    SciTech Connect (OSTI)

    McCoy, J.C.

    1998-07-15

    This safety evaluation for packaging (SEP) evaluates and documents the safe onsite transport of eight high-efficiency particulate air (HEPA) filters in the Pacific Northwest National Laboratory HEPA Filter Box from the 300 Area of the Hanford Site to the Central Waste Complex and on to burial in the 200 West Area. Use of this SEP is authorized for 1 year from the date of release.

  19. Cleanup Verification Package for the 118-F-3, Minor Construction Burial Ground

    SciTech Connect (OSTI)

    M. J. Appel

    2007-01-04

    This cleanup verification package documents completion of remedial action for the 118-F-3, Minor Construction Burial Ground waste site. This site was an open field covered with cobbles, with no vegetation growing on the surface. The site received irradiated reactor parts that were removed during conversion of the 105-F Reactor from the Liquid 3X to the Ball 3X Project safety systems and received mostly vertical safety rod thimbles and step plugs.

  20. Environmental assessment, finding of no significant impact, and response to comments. Radioactive waste storage

    SciTech Connect (OSTI)

    1996-04-01

    The Department of Energy`s (DOE) Rocky Flats Environmental Technology Site (the Site), formerly known as the Rocky Flats Plant, has generated radioactive, hazardous, and mixed waste (waste with both radioactive and hazardous constituents) since it began operations in 1952. Such wastes were the byproducts of the Site`s original mission to produce nuclear weapons components. Since 1989, when weapons component production ceased, waste has been generated as a result of the Site`s new mission of environmental restoration and deactivation, decontamination and decommissioning (D&D) of buildings. It is anticipated that the existing onsite waste storage capacity, which meets the criteria for low-level waste (LL), low-level mixed waste (LLM), transuranic (TRU) waste, and TRU mixed waste (TRUM) would be completely filled in early 1997. At that time, either waste generating activities must cease, waste must be shipped offsite, or new waste storage capacity must be developed.

  1. Filling the Strategic Petroleum Reserve | Department of Energy

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

    Filling the Strategic Petroleum Reserve Filling the Strategic Petroleum Reserve Established in 1975 in the aftermath of the OPEC oil embargo, the Strategic Petroleum Reserve was originally intended to hold at least 750 million barrels of crude oil as an insurance policy against future supply cutoffs (the maximum size was later reduced when a geologically unstable storage site was decommissioned). Today's design capacity is 714 million barrels. Direct Purchases Early fill of the SPR was primarily

  2. Training package 1 for slitting data analysis

    SciTech Connect (OSTI)

    Prime, Michael Bruce

    2015-03-23

    This document and accompanying files are intended as a first training package on how to analyze slitting data. The end goal is to have Idaho National Laboratory (INL) personnel trained to analyze future slitting data taken in the INL Hot Cell on clad, Low-Enriched Uranium (LEU) fuel plates. This first data package will cover data analysis for a monolithic material (as compared to a layered material like the clad fuel plates). The additional issues for layered specimens will be covered in a future training package.

  3. STRUCTURAL AND MECHANICAL EFFECTS OF RADIATION EXPOSURE ON FILLED...

    Office of Scientific and Technical Information (OSTI)

    EFFECTS OF RADIATION EXPOSURE ON FILLED ELASTOMERS - RECENT INSIGHTS FROM EXPERIMENTS AND MODELING Citation Details In-Document Search Title: STRUCTURAL AND MECHANICAL EFFECTS OF ...

  4. Casting Process Simulator 2D Mold Fill and Solidification

    Energy Science and Technology Software Center (OSTI)

    1995-05-06

    The CaPS software is a tool used to setup, simulate, and examine the results from two-dimensional filling and solidification of a sand casting.

  5. EPA - Section 404 Dredge and Fill Permitting webpage | Open Energy...

    Open Energy Info (EERE)

    the discharge of dredged or fill material into waters of the United States. Author Environmental Protection Agency Published Environmental Protection Agency, Date Not...

  6. Theoretical study of Ag- and Au-filled skutterudites.

    Broader source: Energy.gov [DOE]

    Uses ab initio atomistic DFT modeling as implemented in VASP to determine theoretical values of thermoelectric properties for Ag-filled skutterudites.

  7. Oregon Removal-Fill Permit Application Webpage | Open Energy...

    Open Energy Info (EERE)

    Application Webpage Jump to: navigation, search OpenEI Reference LibraryAdd to library Web Site: Oregon Removal-Fill Permit Application Webpage Abstract Provides information for...

  8. Oregon Section 401 Removal/Fill Certification Webpage | Open...

    Open Energy Info (EERE)

    Certification Webpage Jump to: navigation, search OpenEI Reference LibraryAdd to library Web Site: Oregon Section 401 RemovalFill Certification Webpage Abstract Provides overview...

  9. Casting Process Simulator 3D Mold Fill an Solidification

    Energy Science and Technology Software Center (OSTI)

    1997-03-06

    The CAPS software is a tool used to setup, simulate, and examine the results from three-dimensional filling and solidification of a sand casting.

  10. Analysis on Current Status of the Gas Filling Station Networks...

    Open Energy Info (EERE)

    Name: Analysis on Current Status of the Gas Filling Station Networks Website Focus Area: Natural Gas Topics: Potentials & Scenarios Website: www.gashighway.net...

  11. Idaho Dredge and Fill Permits Webpage | Open Energy Information

    Open Energy Info (EERE)

    Not Provided DOI Not Provided Check for DOI availability: http:crossref.org Online Internet link for Idaho Dredge and Fill Permits Webpage Citation Idaho Department of...

  12. The Model 9977 Radioactive Material Packaging Primer (Technical...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: The Model 9977 Radioactive Material Packaging Primer Citation Details In-Document Search Title: The Model 9977 Radioactive Material Packaging Primer The Model...

  13. The Use of Humidity Sensors to Develop BIPV Packaging Solutions...

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

    The Use of Humidity Sensors to Develop BIPV Packaging Solutions The Use of Humidity Sensors to Develop BIPV Packaging Solutions Presented at the PV Module Reliability Workshop, ...

  14. Advanced Framing Systems and Packages - Building America Top...

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

    Advanced Framing Systems and Packages - Building America Top Innovation Advanced Framing Systems and Packages - Building America Top Innovation This photo shows advanced framing ...

  15. Hail Impact Testing on Crystalline Si Modules with Flexible Packaging...

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

    Hail Impact Testing on Crystalline Si Modules with Flexible Packaging Hail Impact Testing on Crystalline Si Modules with Flexible Packaging Presented at the PV Module Reliability ...

  16. Prediction of packaging seal life using thermoanalytical techniques...

    Office of Scientific and Technical Information (OSTI)

    Conference: Prediction of packaging seal life using thermoanalytical techniques Citation Details In-Document Search Title: Prediction of packaging seal life using thermoanalytical ...

  17. CBEI: Packaged Masonry Wall Retrofit Solution for Small and Medium...

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

    Packaged Masonry Wall Retrofit Solution for Small and Medium Sized Commercial Buildings - 2015 Peer Review CBEI: Packaged Masonry Wall Retrofit Solution for Small and Medium Sized ...

  18. White LED with High Package Extraction Efficiency (Technical...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: White LED with High Package Extraction Efficiency Citation Details In-Document Search Title: White LED with High Package Extraction Efficiency The goal of this ...

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

    Office of Environmental Management (EM)

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

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

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

    Ceramic Materials and Packaging Technologies for Realizing Sensors for Concentrating Solar Power Systems Advanced Ceramic Materials and Packaging Technologies for Realizing Sensors ...

  1. Communication Is Key to Packaging and Transportation Safety and...

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

    Communication Is Key to Packaging and Transportation Safety and Compliance Communication Is Key to Packaging and Transportation Safety and Compliance Presentation made by Steve ...

  2. Low-Cost Packaged CHP System with Reduced Emissions - Presentation...

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

    Low-Cost Packaged CHP System with Reduced Emissions - Presentation by Cummins Power Generation, June 2011 Low-Cost Packaged CHP System with Reduced Emissions - Presentation by ...

  3. 2014-08-28 Issuance: Energy Conservation Standards for Packaged...

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

    Air Conditioners and Packaged Terminal Heat Pumps; Notice of Proposed Rulemaking and Public Meeting 2014-08-28 Issuance: Energy Conservation Standards for Packaged Terminal Air ...

  4. The Waste Isolation Pilot Plant Hazardous Waste Facility Permit...

    Office of Environmental Management (EM)

    The Waste Isolation Pilot Plant Hazardous Waste Facility Permit, Waste Analysis Plan The Waste Isolation Pilot Plant Hazardous Waste Facility Permit, Waste Analysis Plan This ...

  5. Simulation of impact of the Generic Accident-Resistant Packaging (GAP)

    SciTech Connect (OSTI)

    Slavin, A.M.

    1994-10-01

    Finite element simulations modelling impact of the Generic Accident-Resistant Packaging (GAP) have been performed. The GAP is a nuclear weapon shipping container that will be used by accident response groups from both the United States and the United Kingdom. The package is a thin-walled steel structure filled with rigid polyurethane foam and weighs approximately 5100 lbs when loaded. The simulations examined 250 ft/s impacts onto a rigid target at several orientations. The development of the finite element model included studies of modelling assumptions and material parameters. Upon completion of the simulation series, three full-scale impact tests were performed. A comparison of the simulation results to the test data is given. Differences between the results and data are examined, and possible explanations for the differences are discussed.

  6. Final Hanford Site Transuranic (TRU) Waste Characterization QA Project Plan

    SciTech Connect (OSTI)

    GREAGER, T.M.

    2000-12-06

    The Quality Assurance Project Plan (QAPjP) has been prepared for waste characterization activities to be conducted by the Transuranic (TRU) Project at the Hanford Site to meet requirements set forth in the Waste Isolation Pilot Plan (WIPP) Hazardous Waste Facility Permit, 4890139088-TSDF, Attachment B, including Attachments B1 through B6 (WAP) (DOE, 1999a). The QAPjP describes the waste characterization requirements and includes test methods, details of planned waste sampling and analysis, and a description of the waste characterization and verification process. In addition, the QAPjP includes a description of the quality assurance/quality control (QA/QC) requirements for the waste characterization program. Before TRU waste is shipped to the WIPP site by the TRU Project, all applicable requirements of the QAPjP shall be implemented. Additional requirements necessary for transportation to waste disposal at WIPP can be found in the ''Quality Assurance Program Document'' (DOE 1999b) and HNF-2600, ''Hanford Site Transuranic Waste Certification Plan.'' TRU mixed waste contains both TRU radioactive and hazardous components, as defined in the WLPP-WAP. The waste is designated and separately packaged as either contact-handled (CH) or remote-handled (RH), based on the radiological dose rate at the surface of the waste container. RH TRU wastes are not currently shipped to the WIPP facility.

  7. Departmental Materials Transportation and Packaging Management

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

    1995-09-27

    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.

  8. Departmental Materials Transportation and Packaging Management

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

    1995-10-26

    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.

  9. Pump packages for Colombian crude oil pipeline

    SciTech Connect (OSTI)

    1994-05-01

    The Caterpillar Large Engine Center recently packaged ten engine-driven centrifugal pump packages for British Petroleum Exploration`s crude oil pipeline in South America. The ten sets, which use Ingersoll-Dresser centrifugal pumps, are designed to increase significantly the output of BP`s Central LLanos pipeline located in a remote region near Bogota, Colombia. BP anticipates that the addition of the new pump packages will increase daily volume from the current 100000 barrels to approximately 210000 barrels when the upgrade of the pipeline is completed in September. The ten sets are installed at three separate pumping stations. The stations are designed to operate continuously while unmanned, with only periodic maintenance required. The pump packages are powered by Caterpillar 3612 engines rated 3040 kW at 1000 r/min. The 12-cylinder engines are turbocharged and charge-air cooled and use the pipeline oil as both fuel and a cooling medium for the fuel injectors.

  10. US Department of Transportation specification packages evaluation

    SciTech Connect (OSTI)

    Ratledge, J.E.; Rawl, R.R.

    1992-01-01

    Specification packages are broad families of package designs and approved by the Department of Transportation (DOT) for transport of certain classes of radioactive materials, with each specification containing a number of designs of various sizes. Many of the individual package designs are not supported by reasonably current safety analyses. The Nuclear Regulatory Commission (NRC) asked Oak Ridge National Laboratory (ORNL) staff to collect all related information, perform analyses, and identify alternative actions that will enable NRC and DOT to make informed decisions on whether to retain, withdraw, or modify the existing regulatory permission for the use of specification packages to transport radioactive and fissile materials. This paper presents the background, issues, and progress made in this activity.

  11. US Department of Transportation specification packages evaluation

    SciTech Connect (OSTI)

    Ratledge, J.E.; Rawl, R.R.

    1992-03-01

    Specification packages are broad families of package designs and approved by the Department of Transportation (DOT) for transport of certain classes of radioactive materials, with each specification containing a number of designs of various sizes. Many of the individual package designs are not supported by reasonably current safety analyses. The Nuclear Regulatory Commission (NRC) asked Oak Ridge National Laboratory (ORNL) staff to collect all related information, perform analyses, and identify alternative actions that will enable NRC and DOT to make informed decisions on whether to retain, withdraw, or modify the existing regulatory permission for the use of specification packages to transport radioactive and fissile materials. This paper presents the background, issues, and progress made in this activity.

  12. Hazardous Materials Packaging and Transportation Safety

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

    2015-04-20

    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.

  13. Shockley-Read-Hall recombination in pre-filled and photo-filled intermediate band solar cells

    SciTech Connect (OSTI)

    Mayani, Maryam Gholami; Reenaas, Turid Worren

    2014-08-18

    In this work, we study how Shockley-Read-Hall (SRH) recombination via energy levels in the bandgap, caused by defects or impurities, affects the performance of both photo-filled and pre-filled intermediate band solar cells (IBSCs). For a pre-filled cell, the IB is half-filled in equilibrium, while it is empty for the photo-filled cell in equilibrium. The energy level, density, and capture cross-sections of the defects/impurities are varied systematically. We find that the photo-filled cells are, in general, less efficient than pre-filled cells, except when the defect level is between the conduction band and the IB. In that case, for a range of light intensities, the photo-filled cell performs better than the pre-filled. When the defect level is at the same energy as the IB, the efficiency is above 82% of the defect-free case, when less than 50% of the states at the IB lead to SRH recombination. This shows that even if SRH recombination via the IB takes place, high efficiencies can be achieved. We also show that band gap optimization can be used to reduce the SRH recombination.

  14. Bioelectrochemical Integration of Waste Heat Recovery, Waste...

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

    Bioelectrochemical Integration of Waste Heat Recovery, Waste-to-Energy Conversion, and Waste-to-Chemical Conversion with Industrial Gas and Chemical Manufacturing Processes ...

  15. HOPSPACK: Hybrid Optimization Parallel Search Package.

    SciTech Connect (OSTI)

    Gray, Genetha A.; Kolda, Tamara G.; Griffin, Joshua; Taddy, Matt; Martinez-Canales, Monica

    2008-12-01

    In this paper, we describe the technical details of HOPSPACK (Hybrid Optimization Parallel SearchPackage), a new software platform which facilitates combining multiple optimization routines into asingle, tightly-coupled, hybrid algorithm that supports parallel function evaluations. The frameworkis designed such that existing optimization source code can be easily incorporated with minimalcode modification. By maintaining the integrity of each individual solver, the strengths and codesophistication of the original optimization package are retained and exploited.4

  16. Packaging material for thin film lithium batteries

    DOE Patents [OSTI]

    Bates, John B.; Dudney, Nancy J.; Weatherspoon, Kim A.

    1996-01-01

    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.

  17. FAQS Qualification Card - NNSA Package Certification Engineer |

    Energy Savers [EERE]

    Department of Energy NNSA Package Certification Engineer FAQS Qualification Card - NNSA Package Certification Engineer A key element for the Department's Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety

  18. Isorropia Partitioning and Load Balancing Package

    Energy Science and Technology Software Center (OSTI)

    2006-09-01

    Isorropia is a partitioning and load balancing package which interfaces with the Zoltan library. Isorropia can accept input objects such as matrices and matrix-graphs, and repartition/redistribute them into a better data distribution on parallel computers. Isorropia is primarily an interface package, utilizing graph and hypergraph partitioning algorithms that are in the Zoltan library which is a third-party library to Tilinos.

  19. Defense High Level Waste Disposal Container System Description

    SciTech Connect (OSTI)

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials will be selected for the disposal container inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lids will be a barrier made of high-nickel alloy. The defense HLW disposal container interfaces with the emplacement drift environment and the internal waste by transferring heat from the canisters to the external environment and by protecting the canisters and their contents from damage/degradation by the external environment. The disposal container also interfaces with the canisters by limiting access of moderator and oxidizing agents to the waste. A loaded and sealed disposal container (waste package) interfaces with the Emplacement Drift System's emplacement drift waste package supports upon which the waste packages are placed. The disposal container interfaces with the Canister Transfer System, Waste Emplacement /Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement, and retrieval for the disposal container/waste package.

  20. Protection of microelectronic devices during packaging

    DOE Patents [OSTI]

    Peterson, Kenneth A.; Conley, William R.

    2002-01-01

    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.

  1. TECHNICAL EVALUATION OF THE SAFE TRANSPORTATION OF WASTE CONTAINERS COATED WITH POLYUREA

    SciTech Connect (OSTI)

    VAIL, T.S.

    2007-03-30

    This technical report is to evaluate and establish that the transportation of waste containers (e.g. drums, wooden boxes, fiberglass-reinforced plywood (FRP) or metal boxes, tanks, casks, or other containers) that have an external application of polyurea coating between facilities on the Hanford Site can be achieved with a level of onsite safety equivalent to that achieved offsite. Utilizing the parameters, requirements, limitations, and controls described in the DOE/RL-2001-36, ''Hanford Sitewide Transportation Safety Document'' (TSD) and the Department of Energy Richland Operations (DOE-RL) approved package specific authorizations (e.g. Package Specific Safety Documents (PSSDs), One-Time Requests for Shipment (OTRSs), and Special Packaging Authorizations (SPAS)), this evaluation concludes that polyurea coatings on packages does not impose an undue hazard for normal and accident conditions. The transportation of all packages on the Hanford Site must comply with the transportation safety basis documents for that packaging system. Compliance with the requirements, limitations, or controls described in the safety basis for a package system will not be relaxed or modified because of the application of polyurea. The inspection criteria described in facility/projects procedures and work packages that ensure compliance with Container Management Programs and transportation safety basis documentation dictate the need to overpack a package without consideration for polyurea. This technical report reviews the transportation of waste packages coated with polyurea and does not credit the polyurea with enhancing the structural, thermal, containment, shielding, criticality, or gas generating posture of a package. Facilities/Projects Container Management Programs must determine if a container requires an overpack prior to the polyurea application recognizing that circumstances newly discovered surface contamination or loss of integrity may require a previously un-overpacked package to subsequently require overpacking. Therefore, the polyurea coating can not be credited to avoid the need to overpack a package or enhance the transportation safety of a structurally sound package that has polyurea on the exterior.

  2. Solar collector panels having coated fibrous filling for fire inhibition

    SciTech Connect (OSTI)

    Reinert, C.P.

    1982-04-13

    Solar collector panels filled with porous fiber mats have the fibers coated with a pigmented intumescent paint which expands to partially fill the spaces between the fibers for retarding convective fluid flow through the fiber mat in the case of a fire in the structure with which the collector is associated.

  3. Filling and solidification with coupled heat transfer and stress analysis

    SciTech Connect (OSTI)

    Ruiz, D.J.; Khandhia, Y.

    1995-12-31

    A full filling and solidification simulation of the MCWASP 7 casting has been carried out using the commercial finite element (FE) analysis codes LS-DYNA3D and FIDAP. The two codes were used to carry out back-to-back analysis of the mould filling. Both codes track the free surface and predict filling times. LS-DYNA3D has been used both for the filling simulation and to carry out a coupled thermal and stress analysis of the casting during solidification, predicting cooling rates, residual stresses and as-cast shape. The methodology used and details of the FE models are summarized. Results are presented for comparison with the casting trial data (x-rays of filling and thermocouple data).

  4. WRAP Module 1 sampling strategy and waste characterization alternatives study

    SciTech Connect (OSTI)

    Bergeson, C.L.

    1994-09-30

    The Waste Receiving and Processing Module 1 Facility is designed to examine, process, certify, and ship drums and boxes of solid wastes that have a surface dose equivalent of less than 200 mrem/h. These wastes will include low-level and transuranic wastes that are retrievably stored in the 200 Area burial grounds and facilities in addition to newly generated wastes. Certification of retrievably stored wastes processing in WRAP 1 is required to meet the waste acceptance criteria for onsite treatment and disposal of low-level waste and mixed low-level waste and the Waste Isolation Pilot Plant Waste Acceptance Criteria for the disposal of TRU waste. In addition, these wastes will need to be certified for packaging in TRUPACT-II shipping containers. Characterization of the retrievably stored waste is needed to support the certification process. Characterization data will be obtained from historical records, process knowledge, nondestructive examination nondestructive assay, visual inspection of the waste, head-gas sampling, and analysis of samples taken from the waste containers. Sample characterization refers to the method or methods that are used to test waste samples for specific analytes. The focus of this study is the sample characterization needed to accurately identify the hazardous and radioactive constituents present in the retrieved wastes that will be processed in WRAP 1. In addition, some sampling and characterization will be required to support NDA calculations and to provide an over-check for the characterization of newly generated wastes. This study results in the baseline definition of WRAP 1 sampling and analysis requirements and identifies alternative methods to meet these requirements in an efficient and economical manner.

  5. CHALLENGES WITH RETRIEVING TRANSURANIC WASTE FROM THE HANFORD BURIAL GROUNDS

    SciTech Connect (OSTI)

    SWAN, R.J.; LAKES, M.E.

    2007-08-06

    The U.S. DOE's Hanford Reservation produced plutonium and other nuclear materials for the nation's defense starting in World War II. The defense mission generated wastes that were either retrievably stored (i.e. retrievably stored waste) and/or disposed of in burial grounds. Challenges have emerged from retrieving suspect TRU waste including adequacy of records, radiological concerns, container integrity, industrial hygiene and safety issues, the lack of processing/treatment facilities, and the integration of regulatory requirements. All retrievably stored waste is managed as mixed waste and assumed to be TRU waste, unless documented otherwise. Mixed waste is defined as radioactive waste that contains hazardous constituents. The Atomic Energy Act governs waste with radionuclides, and the Resource Conservation and Recovery Act (RCRA) governs waste with hazardous constituents. Waste may also be governed by the Toxic Substances Control Act (TSCA), and a portion may be managed under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). In 1970, TRU waste was required to be placed in 20-year retrievable storage and segregated from other Waste. Prior to that date, segregation did not occur. Because of the changing definition of TRU over the years, and the limitations of early assay equipment, all retrievably stored waste in the burial grounds is managed as suspect TRU. Experience has shown that some of this waste will be characterized as low-level (non-TRU) waste after assay. The majority of the retrieved waste is not amenable to sampling due to waste type and/or radiological issues. Key to waste retrieval and disposition are characterization, historical investigation and research, knowledge of past handling and packaging, as well as a broad understanding and application of the regulations.

  6. Example Performance Targets and Efficiency Packages Greensburg, Kansas (Presentation)

    SciTech Connect (OSTI)

    Anderson, R.

    2008-01-01

    This presentation shows the energy performance targets and efficiency packages for residential buildings in Greensburg, Kansas.

  7. Filling Knowledge Gaps with Five Fuel Cycle Studies

    SciTech Connect (OSTI)

    Steven J. Piet; Jess Gehin; William Halsey; Temitope Taiwo

    2010-11-01

    During FY 2010, five studies were conducted of technology families’ applicability to various fuel cycle strategies to fill in knowledge gaps in option space and to better understand trends and patterns. Here, a “technology family” is considered to be defined by a type of reactor and by selection of which actinides provide fuel. This report summarizes the higher-level findings; the detailed analyses and results are documented in five individual reports, as follows: • Advanced once through with uranium fuel in fast reactors (SFR), • Advanced once through (uranium fuel) or single recycle (TRU fuel) in high temperature gas cooled reactors (HTGR), • Sustained recycle with Th/U-233 in light water reactors (LWRs), • Sustained recycle with Th/U-233 in molten salt reactors (MSR), and • Several fuel cycle missions with Fusion-Fission Hybrid (FFH). Each study examined how the designated technology family could serve one or more designated fuel cycle missions, filling in gaps in overall option space. Each study contains one or more illustrative cases that show how the technology family could be used to meet a fuel cycle mission, as well as broader information on the technology family such as other potential fuel cycle missions for which insufficient information was available to include with an illustrative case. None of the illustrative cases can be considered as a reference, baseline, or nominal set of parameters for judging performance; the assessments were designed to assess areas of option space and were not meant to be optimized. There is no implication that any of the cases or technology families are necessarily the best way to meet a given fuel cycle mission. The studies provide five examples of 1-year fuel cycle assessments of technology families. There is reasonable coverage in the five studies of the performance areas of waste management and uranium utilization. The coverage of economics, safety, and proliferation resistance and physical protection in the five studies was spotty. Some studies did not have existing or past work to draw on in one or more of these areas. Resource constraints limited the amount of new analyses that could be performed. Little or no assessment was done of how soon any of the technologies could be deployed and therefore how quickly they could impact domestic or international fuel cycle performance. There were six common R&D needs, such as the value of advanced fuels, cladding, coating, and structure that would survive high neutron fluence. When a technology family is considered for use in a new fuel cycle mission, fuel cycle performance characteristics are dependent on both the design choices and the fuel cycle approach. For example, the use of the sodium-cooled fast reactor to provide recycle in either breeder or burner mode has been studied for decades, but the SFR could be considered for once-through fuel cycle with the physical reactor design and fuel management parameters changed. In addition, the sustained recycle with Th/U-233 in LWR could be achieved with a heterogeneous assembly and derated power density. Therefore, it may or may not be adjustable for other fuel cycle missions although a reactor intended for one fuel cycle mission is built. Simple parameter adjustment in applying a technology family to a new fuel cycle mission should be avoided and, if observed, the results viewed with caution.

  8. Estimating Waste Inventory and Waste Tank Characterization

    Broader source: Energy.gov [DOE]

    Summary Notes from 28 May 2008 Generic Technical Issue Discussion on Estimating Waste Inventory and Waste Tank Characterization

  9. Plutonium-238 Transuranic Waste Decision Analysis

    SciTech Connect (OSTI)

    Brown, Mike; Lechel, David J.; Leigh, C.D.

    1999-06-29

    Five transuranic (TRU) waste sites in the Department of Energy (DOE) complex, collectively, have more than 2,100 cubic meters of Plutonium-238 (Pu-238) TRU waste that exceed the wattage restrictions of the Transuranic Package Transporter-II (TRUPACT-11). The Waste Isolation Pilot Plant (WIPP) is being developed by the DOE as a repository for TRU waste. With the Waste Isolation Pilot Plant (WIPP) opening in 1999, these sites are faced with a need to develop waste management practices that will enable the transportation of Pu-238 TRU waste to WIPP for disposal. This paper describes a decision analysis that provided a logical framework for addressing the Pu-238 TRU waste issue. The insights that can be gained by performing a formalized decision analysis are multifold. First and foremost, the very process. of formulating a decision tree forces the decision maker into structured, logical thinking where alternatives can be evaluated one against the other using a uniform set of criteria. In the process of developing the decision tree for transportation of Pu-238 TRU waste, several alternatives were eliminated and the logical order for decision making was discovered. Moreover, the key areas of uncertainty for proposed alternatives were identified and quantified. The decision analysis showed that the DOE can employ a combination approach where they will (1) use headspace gas analyses to show that a fraction of the Pu-238 TRU waste drums are no longer generating hydrogen gas and can be shipped to WIPP ''as-is'', (2) use drums and bags with advanced filter systems to repackage Pu-238 TRU waste drums that are still generating hydrogen, and (3) add hydrogen getter materials to the inner containment vessel of the TRUPACT-11to relieve the build-up of hydrogen gas during transportation of the Pu-238 TRU waste drums.

  10. Waste handling activities in glovebox dismantling facility

    SciTech Connect (OSTI)

    Kitamura, Akihiro; Okada, Takashi; Kashiro, Kashio; Yoshino, Masanori; Hirano, Hiroshi

    2007-07-01

    The Glovebox Dismantling Facility is a facility to decontaminate and size-reduce after-service gloveboxes in the Plutonium Fuel Production Facility, Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency. The wastes generated from these dismantling activities are simultaneously handled and packaged into drums in a bag-out manner. For future waste treatment and disposal, these wastes are separated into material categories. In this paper, we present the basic steps and analyzed data for the waste handling activities. The data were collected from dismantling activities for three gloveboxes (Grinding Pellet Glovebox, Visual Inspection Glovebox, Outer-diameter Screening Glovebox) conducted from 2001-2004. We also describe both current and near-future improvements. (authors)

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

    SciTech Connect (OSTI)

    1996-10-01

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

  12. Paducah Package Steam Boilers to Provide Efficiency, Environmental Benefits

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

    | Department of Energy Package Steam Boilers to Provide Efficiency, Environmental Benefits Paducah Package Steam Boilers to Provide Efficiency, Environmental Benefits October 29, 2015 - 12:10pm Addthis An aerial view of the package boilers installed into the siteĆ¢Ā€Ā™s existing steam system. An aerial view of the package boilers installed into the site's existing steam system. Pipefitters Mike Askren, left, and Ron Parrot install the water inlet on one of the package boilers. Pipefitters

  13. Disposal of Rocky Flats residues as waste

    SciTech Connect (OSTI)

    Dustin, D.F.; Sendelweck, V.S. . Rocky Flats Plant); Rivera, M.A. )

    1993-01-01

    Work is underway at the Rocky Flats Plant to evaluate alternatives for the removal of a large inventory of plutonium-contaminated residues from the plant. One alternative under consideration is to package the residues as transuranic wastes for ultimate shipment to the Waste Isolation Pilot Plant. Current waste acceptance criteria and transportation regulations require that approximately 1000 cubic yards of residues be repackaged to produce over 20,000 cubic yards of WIPP certified waste. The major regulatory drivers leading to this increase in waste volume are the fissile gram equivalent, surface radiation dose rate, and thermal power limits. In the interest of waste minimization, analyses have been conducted to determine, for each residue type, the controlling criterion leading to the volume increase, the impact of relaxing that criterion on subsequent waste volume, and the means by which rules changes may be implemented. The results of this study have identified the most appropriate changes to be proposed in regulatory requirements in order to minimize the costs of disposing of Rocky Flats residues as transuranic wastes.

  14. Disposal of Rocky Flats residues as waste

    SciTech Connect (OSTI)

    Dustin, D.F.; Sendelweck, V.S.; Rivera, M.A.

    1993-03-01

    Work is underway at the Rocky Flats Plant to evaluate alternatives for the removal of a large inventory of plutonium-contaminated residues from the plant. One alternative under consideration is to package the residues as transuranic wastes for ultimate shipment to the Waste Isolation Pilot Plant. Current waste acceptance criteria and transportation regulations require that approximately 1000 cubic yards of residues be repackaged to produce over 20,000 cubic yards of WIPP certified waste. The major regulatory drivers leading to this increase in waste volume are the fissile gram equivalent, surface radiation dose rate, and thermal power limits. In the interest of waste minimization, analyses have been conducted to determine, for each residue type, the controlling criterion leading to the volume increase, the impact of relaxing that criterion on subsequent waste volume, and the means by which rules changes may be implemented. The results of this study have identified the most appropriate changes to be proposed in regulatory requirements in order to minimize the costs of disposing of Rocky Flats residues as transuranic wastes.

  15. Tanks 18 And 19-F Structural Flowable Grout Fill Material Evaluation And Recommendations

    SciTech Connect (OSTI)

    Langton, C. A.; Stefanko, D. B.

    2013-04-23

    Cementitious grout will be used to close Tanks 18-F and 19-F. The functions of the grout are to: 1) physically stabilize the final landfill by filling the empty volume in the tanks with a non-compressible material; 2) provide a barrier for inadvertent intrusion into the tank; 3) reduce contaminant mobility by a) limiting the hydraulic conductivity of the closed tank and b) reducing contact between the residual waste and infiltrating water; and 4) providing an alkaline, chemically reducing environment in the closed tank to control speciation and solubility of selected radionuclides. The objective of this work was to identify a single (all-in-one) grout to stabilize and isolate the residual radionuclides in the tank, provide structural stability of the closed tank and serve as an inadvertent intruder barrier. This work was requested by V. A. Chander, High Level Waste (HLW) Tank Engineering, in HLW-TTR-2011-008. The complete task scope is provided in the Task Technical and QA Plan, SRNL-RP-2011-00587 Revision 0. The specific objectives of this task were to: 1) Identify new admixtures and dosages for formulating a zero bleed flowable tank fill material selected by HLW Tank Closure Project personnel based on earlier tank fill studies performed in 2007. The chemical admixtures used for adjusting the flow properties needed to be updated because the original admixture products are no longer available. Also, the sources of cement and fly ash have changed, and Portland cements currently available contain up to 5 wt. % limestone (calcium carbonate). 2) Prepare and evaluate the placement, compressive strength, and thermal properties of the selected formulation with new admixture dosages. 3) Identify opportunities for improving the mix selected by HLW Closure Project personnel and prepare and evaluate two potentially improved zero bleed flowable fill design concepts; one based on the reactor fill grout and the other based on a shrinkage compensating flowable fill mix design. 4) Prepare samples for hydraulic property measurements for comparison to the values in the F and H- Tank Farm Performance Assessments (PAs). 5) Identify a grout mix for the Tanks 18-F and 19-F Grout Procurement Specification [Forty, 2011 a, b, c]. Results for two flowable zero bleed structural fill concepts containing 3/8 inch gravel (70070 Series and LP#8 Series) and a sand only mix (SO Series) are provided in this report. Tank Farm Engineering and SRNL Project Management selected the 70070 mix as the base case for inclusion in Revision 0 of the Tanks 18-F and 19-F grout procurement specification [Forty 2011 a] and requested admixture recommendations and property confirmation for this formulation [Forty, 2011 b]. Lower cementitious paste mixes were formulated because the 70070 mix is over designed with respect to strength and generates more heat from hydration reactions than is desirable for mass pour application. Work was also initiated on a modification of the recommended mix which included shrinkage compensation to mitigate fast pathways caused by shrinkage cracking and poor physical bonding to the tank and ancillary equipment. Testing of this option was postponed to FY12. Mix, LP#8-16 is recommended for inclusion in the specification for furnishing and delivering tank closure grout for Tanks 18-F and 19-F [Forty, 2011 c]. A shrinkage compensating variation of this mix, LP#16C, has not been fully developed and characterized at this time.

  16. TANKS 18 AND 19-F STRUCTURAL FLOWABLE GROUT FILL MATERIAL EVALUATION AND RECOMMENDATIONS

    SciTech Connect (OSTI)

    Stefanko, D.; Langton, C.

    2011-11-01

    Cementitious grout will be used to close Tanks 18-F and 19-F. The functions of the grout are to: (1) physically stabilize the final landfill by filling the empty volume in the tanks with a non compressible material; (2) provide a barrier for inadvertent intrusion into the tank; (3) reduce contaminant mobility by (a) limiting the hydraulic conductivity of the closed tank and (b) reducing contact between the residual waste and infiltrating water; and (4) providing an alkaline, chemically reducing environment in the closed tank to control speciation and solubility of selected radionuclides. The objective of this work was to identify a single (all-in-one) grout to stabilize and isolate the residual radionuclides in the tank, provide structural stability of the closed tank and serve as an inadvertent intruder barrier. This work was requested by V. A. Chander, High Level Waste (HLW) Tank Engineering, in HLW-TTR-2011-008. The complete task scope is provided in the Task Technical and QA Plan, SRNL-RP-2011-00587 Revision 0. The specific objectives of this task were to: (1) Identify new admixtures and dosages for formulating a zero bleed flowable tank fill material selected by HLW Tank Closure Project personnel based on earlier tank fill studies performed in 2007. The chemical admixtures used for adjusting the flow properties needed to be updated because the original admixture products are no longer available. Also, the sources of cement and fly ash have changed, and Portland cements currently available contain up to 5 wt. % limestone (calcium carbonate). (2) Prepare and evaluate the placement, compressive strength, and thermal properties of the selected formulation with new admixture dosages. (3) Identify opportunities for improving the mix selected by HLW Closure Project personnel and prepare and evaluate two potentially improved zero bleed flowable fill design concepts; one based on the reactor fill grout and the other based on a shrinkage compensating flowable fill mix design. (4) Prepare samples for hydraulic property measurements for comparison to the values in the F and H- Tank Farm Performance Assessments (PAs). (5) Identify a grout mix for the Tanks 18-F and 19-F Grout Procurement Specification [Forty, 2011 a, b, c]. Results for two flowable zero bleed structural fill concepts containing 3/8 inch gravel (70070 Series and LP-8 Series) and a sand only mix (SO Series) are provided in this report. Tank Farm Engineering and SRNL Project Management selected the 70070 mix as the base case for inclusion in Revision 0 of the Tanks 18-F and 19-F grout procurement specification [Forty 2011 a] and requested admixture recommendations and property confirmation for this formulation [Forty, 2011 b]. Lower cementitious paste mixes were formulated because the 70070 mix is over designed with respect to strength and generates more heat from hydration reactions than is desirable for mass pour application. Work was also initiated on a modification of the recommended mix which included shrinkage compensation to mitigate fast pathways caused by shrinkage cracking and poor physical bonding to the tank and ancillary equipment. Testing of this option was postponed to FY12.

  17. Waste Management Plan for the Oak Ridge National Remedial Investigation/Feasibility Study

    SciTech Connect (OSTI)

    Not Available

    1988-04-01

    In accordance with the requirements of the Remedial Investigation/Feasibility Study (RI/FS) Project Quality Assurance Plan, this Waste Management Plan establishes clear lines of responsibility and authority, documentation requirements, and operational guidance for the collection, identification, segregation, classification, packaging, certification, and storage/disposal of wastes. These subjects are discussed in the subsequent sections of this document.

  18. Basalt waste isolation project. Quarterly report, April 1, 1981-June 30, 1981

    SciTech Connect (OSTI)

    Deju, R.A.

    1981-08-01

    This document reports progress made in the Basalt Waste Isolation Project during the third quarter of fiscal year 1981. Efforts are described for the following programs of the project work breakdown structure: systems; waste package; site; repository; regulatory and institutional; test facilities; in situ test facilities.

  19. MEMS Packaging - Current Issues and Approaches

    SciTech Connect (OSTI)

    DRESSENDORFER,PAUL V.; PETERSON,DAVID W.; REBER,CATHLEEN ANN

    2000-01-19

    The assembly and packaging of MEMS (Microelectromechanical Systems) devices raise a number of issues over and above those normally associated with the assembly of standard microelectronic circuits. MEMS components include a variety of sensors, microengines, optical components, and other devices. They often have exposed mechanical structures which during assembly require particulate control, space in the package, non-contact handling procedures, low-stress die attach, precision die placement, unique process schedules, hermetic sealing in controlled environments (including vacuum), and other special constraints. These constraints force changes in the techniques used to separate die on a wafer, in the types of packages which can be used in the assembly processes and materials, and in the sealing environment and process. This paper discusses a number of these issues and provides information on approaches being taken or proposed to address them.

  20. DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Blanton, P.

    2013-10-10

    A new Bulk Tritium Shipping Package (BTSP) was designed by the Savannah River National Laboratory to be a replacement for a package that has been used to ship tritium in a variety of content configurations and forms since the early 1970s. The BTSP was certified by the National Nuclear Safety Administration in 2011 for shipments of up to 150 grams of Tritium. Thirty packages were procured and are being delivered to various DOE sites for operational use. This paper summarizes the design features of the BTSP, as well as associated engineered material improvements. Fabrication challenges encountered during production are discussed as well as fielding requirements. Current approved tritium content forms (gas and tritium hydrides), are reviewed, as well as, a new content, tritium contaminated water on molecular sieves. Issues associated with gas generation will also be discussed.

  1. DEVELOPMENT OF THE H1700 SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Abramczyk, G.; Loftin, B.; Mann, P.

    2009-06-05

    The H1700 Package is based on the DOE-EM Certified 9977 Packaging. The H1700 will be certified by the Packaging Certification Division of the National Nuclear Security Administration for the shipment of plutonium by air by the United Stated Military both within the United States and internationally. The H1700 is designed to ship radioactive contents in assemblies of Radioisotope Thermoelectric Generators (RTGs) or arrangements of nested food-pack cans. The RTG containers are designed and tested to remain leaktight during transport, handling, and storage; however, their ability to remain leaktight during transport in the H1700 is not credited. This paper discusses the design and special operation of the H1700.

  2. Geometric Modeling, Radiation Simulation, Rendering, Analysis Package

    Energy Science and Technology Software Center (OSTI)

    1995-01-17

    RADIANCE is intended to aid lighting designers and architects by predicting the light levels and appearance of a space prior to construction. The package includes programs for modeling and translating scene geometry, luminaire data and material properties, all of which are needed as input to the simulation. The lighting simulation itself uses ray tracing techniques to compute radiance values (ie. the quantity of light passing through a specific point in a specific direction), which aremoreĀ Ā» typically arranged to form a photographic quality image. The resulting image may be analyzed, displayed and manipulated within the package, and converted to other popular image file formats for export to other packages, facilitating the production of hard copy output.Ā«Ā less

  3. CHP Integrated with Burners for Packaged Boilers

    SciTech Connect (OSTI)

    Castaldini, Carlo; Darby, Eric

    2013-09-30

    The objective of this project was to engineer, design, fabricate, and field demonstrate a Boiler Burner Energy System Technology (BBEST) that integrates a low-cost, clean burning, gas-fired simple-cycle (unrecuperated) 100 kWe (net) microturbine (SCMT) with a new ultra low-NOx gas-fired burner (ULNB) into one compact Combined Heat and Power (CHP) product that can be retrofit on new and existing industrial and commercial boilers in place of conventional burners. The Scope of Work for this project was segmented into two principal phases: (Phase I) Hardware development, assembly and pre-test and (Phase II) Field installation and demonstration testing. Phase I was divided into five technical tasks (Task 2 to 6). These tasks covered the engineering, design, fabrication, testing and optimization of each key component of the CHP system principally, ULNB, SCMT, assembly BBEST CHP package, and integrated controls. Phase I work culminated with the laboratory testing of the completed BBEST assembly prior to shipment for field installation and demonstration. Phase II consisted of two remaining technical tasks (Task 7 and 8), which focused on the installation, startup, and field verification tests at a pre-selected industrial plant to document performance and attainment of all project objectives. Technical direction and administration was under the management of CMCE, Inc. Altex Technologies Corporation lead the design, assembly and testing of the system. Field demonstration was supported by Leva Energy, the commercialization firm founded by executives at CMCE and Altex. Leva Energy has applied for patent protection on the BBEST process under the trade name of Power Burner and holds the license for the burner currently used in the product. The commercial term Power Burner is used throughout this report to refer to the BBEST technology proposed for this project. The project was co-funded by the California Energy Commission and the Southern California Gas Company (SCG), a division of Sempra Energy. These match funds were provided via concurrent contracts and investments available via CMCE, Altex, and Leva Energy The project attained all its objectives and is considered a success. CMCE secured the support of GI&E from Italy to supply 100 kW Turbec T-100 microturbines for the project. One was purchased by the projectā€™s subcontractor, Altex, and a second spare was purchased by CMCE under this project. The microturbines were then modified to convert from their original recuperated design to a simple cycle configuration. Replacement low-NOx silo combustors were designed and bench tested in order to achieve compliance with the California Air Resources Board (CARB) 2007 emission limits for NOx and CO when in CHP operation. The converted microturbine was then mated with a low NOx burner provided by Altex via an integration section that allowed flow control and heat recovery to minimize combustion blower requirements; manage burner turndown; and recover waste heat. A new fully integrated control system was designed and developed that allowed one-touch system operation in all three available modes of operation: (1) CHP with both microturbine and burner firing for boiler heat input greater than 2 MMBtu/hr; (2) burner head only (BHO) when the microturbine is under service; and (3) microturbine only when boiler heat input requirements fall below 2 MMBtu/hr. This capability resulted in a burner turndown performance of nearly 10/1, a key advantage for this technology over conventional low NOx burners. Key components were then assembled into a cabinet with additional support systems for generator cooling and fuel supply. System checkout and performance tests were performed in the laboratory. The assembled system and its support equipment were then shipped and installed at a host facility where final performance tests were conducted following efforts to secure fabrication, air, and operating permits. The installed power burner is now in commercial operation and has achieved all the performance goals.

  4. 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.; Scales, Charlie R.; Maddrell, Ewan R.; Hobbs, Jeff

    2013-07-01

    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)

  5. Express Package Shipping Services | The Ames Laboratory

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

    Express Package Shipping Services General Information: The materials handling office's express package shipping service offers overnight domestic letter and parcel service for official business only. Express international shipments are also offered. Many hazardous and non-hazardous materials may be shipped by express. This service is provided to Ames Laboratory personnel. Hours: Monday through Friday - 7:30 a.m. to 11:50 a.m. - 12:30 p.m. through 4:00 p.m. Cutoff: for 10:30 a.m. next day

  6. 1 and 2-Dimensional Line Transfer Package

    Energy Science and Technology Software Center (OSTI)

    1990-07-01

    LXF1D is a one dimensional steady-state line transfer package designed to handle: overlapping and or interacting lines, planar, cylindrical, spherical (and special) geometries, doppler shifts, complete redistribution (CRD), partial redistribution (PRD). PRD requires the use of REDIST or some other package to produce emission profiles. LXF2D is a two dimensional version of LXF1D for xy and rz geometries. Both LXF1D and LXF2D are designed to be added to existing non-local thermodynamic equilibrium (NLTE) codes withmoreĀ Ā» a minimum of effort.Ā«Ā less

  7. MEMS packaging efforts at Sandia National Laboratories.

    SciTech Connect (OSTI)

    Custer, Jonathan Sloane

    2003-02-01

    Sandia National Laboratories has programs covering a broad range of MEMS technologies from LIGA to bulk to surface micromachining. These MEMS technologies are being considered for an equally broad range of applications, including sensors, actuators, optics, and microfluidics. As these technologies have moved from the research to the prototype product stage, packaging has been required to develop new capabilities to integrated MEMS and other technologies into functional microsystems. This paper discusses several of Sandia's MEMS packaging efforts, focusing mainly on inserting Sandia's SUMMIT V (5-level polysilicon) surface micromachining technology into fieldable microsystems.

  8. Safety evaluation for packaging (onsite) SERF cask

    SciTech Connect (OSTI)

    Edwards, W.S.

    1997-10-24

    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.

  9. Energy Savings Measure Packages. Existing Homes

    SciTech Connect (OSTI)

    Casey, Sean; Booten, Chuck

    2011-11-01

    This document presents the most cost effective Energy Savings Measure Packages (ESMP) for existing mixed-fuel and all electric homes to achieve 15% and 30% savings for each BetterBuildings grantee location across the United States. These packages are optimized for minimum cost to homeowners for source energy savings given the local climate and prevalent building characteristics (i.e. foundation types). Maximum cost savings are typically found between 30% and 50% energy savings over the reference home; this typically amounts to $300 - $700/year.

  10. Energy Savings Measure Packages: Existing Homes

    SciTech Connect (OSTI)

    Casey, S.; Booten, C.

    2011-11-01

    This document presents the most cost effective Energy Savings Measure Packages (ESMP) for existing mixed-fuel and all electric homes to achieve 15% and 30% savings for each BetterBuildings grantee location across the US. These packages are optimized for minimum cost to homeowners for given source energy savings given the local climate and prevalent building characteristics (i.e. foundation types). Maximum cost savings are typically found between 30% and 50% energy savings over the reference home. The dollar value of the maximum annual savings varies significantly by location but typically amounts to $300 - $700/year.

  11. New Mexico Surface Water Quality Bureau Federal Dredge and Fill...

    Open Energy Info (EERE)

    Jump to: navigation, search OpenEI Reference LibraryAdd to library Web Site: New Mexico Surface Water Quality Bureau Federal Dredge and Fill Permits webpage Author New Mexico...

  12. Terminal Operations for Tube Trailer and Liquid Tanker Filling...

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

    Tanker Filling: Status, Challenges and R&D Needs 1 Linde covers the entire hydrogen value chain LH 2 storage SupplyStorage CompressionTransfer Dispenser Onsite SMR Onsite ...

  13. Fracture and thermal aging of resin-filled silicone elastomers...

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect Search Results Journal Article: Fracture and thermal aging of resin-filled ... Publication Date: 2016-03-24 OSTI Identifier: 1236865 Resource Type: Journal Article ...

  14. Innovative Way to Test Batteries Fills a Market Niche | NREL

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

    Innovative Way to Test Batteries Fills a Market Niche November 20, 2014 Illustration of a ... safer, longer-lasting, and more cost-effective batteries for in a variety of applications. ...

  15. System and method for filling a plurality of isolated vehicle fluid circuits through a common fluid fill port

    DOE Patents [OSTI]

    Sullivan, Scott C; Fansler, Douglas

    2014-10-14

    A vehicle having multiple isolated fluid circuits configured to be filled through a common fill port includes a first fluid circuit disposed within the vehicle, the first fluid circuit having a first fill port, a second fluid circuit disposed within the vehicle, and a conduit defining a fluid passageway between the first fluid circuit and second fluid circuit, the conduit including a valve. The valve is configured such that the first and second fluid circuits are fluidly coupled via the passageway when the valve is open, and are fluidly isolated when the valve is closed.

  16. Issues and Experiences on Radioactive Waste Quality Control / Quality Assurance with Regard to Future Disposal

    SciTech Connect (OSTI)

    Beckmerhagen, I.; Brennecke, P.; Steyer, S.; Bandt, G.

    2006-07-01

    In the Federal Republic of Germany all types of radioactive waste (short-lived, long-lived) are to be disposed of in deep geological formations. Thus, the safe management of radioactive waste presupposes an appropriate conditioning of primary waste-to-waste packages suitable for emplacement in a repository as well as the documentation of pre-treatment, processing and packaging steps and the waste package characteristics being relevant for disposal. Due to the operation, decommissioning and dismantling of nuclear facilities as well as the application of radioisotopes in industry, medicine and research and development radioactive waste continuously arises in Germany. In order to manage this waste different measures and procedures regarding its conditioning and quality control/quality assurance were introduced and since many years successfully applied. Waste conditioning is especially characterized by a flexible application of the Konrad waste acceptance requirements. The rationale for this approach is due to the present non-availability of a repository in Germany. Several examples of a 'tailor-made' application of the waste acceptance requirements in treatment, conditioning and documentation processes as well as the quality assurance/quality control processes illustrate the current German approach. (authors)

  17. New Contracts Awarded for Continued Fill of Strategic Petroleum Reserve |

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

    Department of Energy Contracts Awarded for Continued Fill of Strategic Petroleum Reserve New Contracts Awarded for Continued Fill of Strategic Petroleum Reserve February 15, 2005 - 10:12am Addthis WASHINGTON, DC -- The U.S. Department of Energy has awarded two new contracts to deliver crude oil to the Strategic Petroleum Reserve (SPR) this spring under the Royalty-In-Kind (RIK) exchange program. Shell Trading (US) Company and Vitol SA Inc. submitted the best offers and were awarded

  18. DOE to Resume Filling Strategic Petroleum Reserve: Oil Acquisition Slated

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

    for 2009 | Department of Energy to Resume Filling Strategic Petroleum Reserve: Oil Acquisition Slated for 2009 DOE to Resume Filling Strategic Petroleum Reserve: Oil Acquisition Slated for 2009 January 2, 2009 - 9:27am Addthis WASHINGTON, DC -- The U.S. Department of Energy today announced that it plans to take advantage of the recent large decline in crude oil prices, and has issued a solicitation to purchase approximately 12 million barrels of crude oil for the nation's Strategic Petroleum

  19. Fluorinated diamond particles bonded in a filled fluorocarbon resin matrix

    DOE Patents [OSTI]

    Taylor, G.W.; Roybal, H.E.

    1983-11-14

    A method of producing fluorinated diamond particles bonded in a filled fluorocarbon resin matrix. Simple hot pressing techniques permit the formation of such matrices from which diamond impregnated grinding tools and other articles of manufacture can be produced. Teflon fluorocarbon resins filled with Al/sub 2/O/sub 3/ yield grinding tools with substantially improved work-to-wear ratios over grinding wheels known in the art.

  20. Fluorinated diamond particles bonded in a filled fluorocarbon resin matrix

    DOE Patents [OSTI]

    Taylor, Gene W.; Roybal, Herman E.

    1985-01-01

    A method of producing fluorinated diamond particles bonded in a filled fluorocarbon resin matrix. Simple hot pressing techniques permit the formation of such matrices from which diamond impregnated grinding tools and other articles of manufacture can be produced. Teflon fluorocarbon resins filled with Al.sub.2 O.sub.3 yield grinding tools with substantially improved work-to-wear ratios over grinding wheels known in the art.

  1. DOE to Resume Filling Strategic Petroleum Reserve | Department of Energy

    Energy Savers [EERE]

    to Resume Filling Strategic Petroleum Reserve DOE to Resume Filling Strategic Petroleum Reserve January 2, 2009 - 12:00pm Addthis Washington, D.C. - The U.S. Department of Energy today announced that it plans to take advantage of the recent large decline in crude oil prices, and has issued a solicitation to purchase approximately 12 million barrels of crude oil for the nation's Strategic Petroleum Reserve (SPR) to replenish SPR supplies sold following hurricanes Katrina and Rita in 2005. In

  2. STRUCTURAL AND MECHANICAL EFFECTS OF RADIATION EXPOSURE ON FILLED

    Office of Scientific and Technical Information (OSTI)

    ELASTOMERS - RECENT INSIGHTS FROM EXPERIMENTS AND MODELING (Conference) | SciTech Connect Conference: STRUCTURAL AND MECHANICAL EFFECTS OF RADIATION EXPOSURE ON FILLED ELASTOMERS - RECENT INSIGHTS FROM EXPERIMENTS AND MODELING Citation Details In-Document Search Title: STRUCTURAL AND MECHANICAL EFFECTS OF RADIATION EXPOSURE ON FILLED ELASTOMERS - RECENT INSIGHTS FROM EXPERIMENTS AND MODELING Authors: Maiti, A ; Weisgraber, T ; Gee, R ; Dinh, L ; Wilson, T ; Small, W ; Alviso, C ; Chinn, S ;

  3. Settling of loose-fill insulations due to vibration

    SciTech Connect (OSTI)

    Yarbrough, D.W.; McElroy, D.L.; Wright, J.W.

    1981-12-01

    Vibration and impact testing of loose-fill cellulosic, fiberglass, and rock wool insulations has been carried out to provide a data base for settled density tests. The ratio of final density to initial density for the three materials has been determined for repeated 19-mm (0.75-in.) drops, repeated 152-mm (6.0-in.) drops, and vibrations at frequencies from 10 to 60 Hz with displacements from 0.1 mm (0.004 in.) to 6.35 mm (0.25 in.). Repeated 19-mm or 152-mm drops increased the density ratio for rock wool insulation specimens the most, while the cellulosic insulation specimens were affected the least. Density ratios after 200 19-mm drops averaged 1.75 for loose-fill rock wool, 1.45 for loose-fill fiberglass, and 1.27 for loose-fill cellulosic insulations. Vibration tests for 7200 s at 0.1-mm displacement and 15 Hertz produced negligible changes in the densities of all three loose-fill insulations. An 1800-s vibration test at 2.5 mm (0.1 in.) and 10 Hz resulted in average density ratios of 1.05, 1.11, and 1.18 for specimens of loose-fill cellulosic, rock wool, and fiberglass insulations, respectively. Changes in either frequency of vibration, displacement, or test duration can be used to achieve a wide range of laboratory results. Efforts to correlate laboratory results with in situ density measurements are presented.

  4. Settling of loose-fill insulations due to vibration

    SciTech Connect (OSTI)

    Yarbrough, D.W.; Wright, J.H.; McElroy, D.L.; Scanlan, T.F.

    1983-01-01

    Vibration and impact testing of loose-fill cellulosic, fiberglass, and rock wool insulations has been carried out to provide a data base for settled density tests. The ratio of final density to initial density for the three materials has been determined for repeated 19-mm (0.75-in.) drops, repeated 152-mm (6.0-in.) drops, and vibrations at frequencies from 10 to 60 Hz with displacements from 0.1 mm (0.004 in.) to 6.35 mm (0.25 in.). Repeated 19-mm or 152-mm drops increased the density ratio for rock wool insulation specimens the most, while the cellulosic insulation specimens were affected the least. Density ratios after 200 19-mm drops averaged 1.75 for loose-fill rock wool, 1.45 for loose-fill fiberglass, and 1.27 for loose-fill cellulosic insulations. Vibration tests for 7200 s at 0.1-mm displacement and 15 Hertz produced negligible changes in the densities of all three loose-fill insulations. An 1800-s vibration test at 2.5 mm (0.1 in.) and 10 Hz resulted in average density ratios of 1.05, 1.11, and 1.18 for specimens of loose-fill cellulosic, rock wool, and fiberglass insulations, respectively. Changes in either frequency of vibration, displacement, or test duration can be used to achieve a wide range of laboratory results. Efforts to correlate laboratory results with in situ density measurements are presented.

  5. Low-level radioactive waste (LLW) management at the Nevada Test Site (NTS)

    SciTech Connect (OSTI)

    Becker, B.D.; Gertz, C.P.; Clayton, W.A.; Crowe, B.M.

    1998-12-31

    In 1978, the Department of Energy, Nevada Operations Office (DOE/NV), established a managed LLW disposal project at the Nevada Test Site (NTS). Two, sites which were already accepting limited amounts of on-site generated waste for disposal and off-site generated Transuranic Waste for interim storage, were selected to house the disposal facilities. In those early days, these sites, located about 15 miles apart, afforded the DOE/NV the opportunity to use at least two technologies to manage its waste cost effectively. The Area 5 Radioactive Waste Management Site (RWMS) uses engineered shallow-land burial cells to dispose packaged waste while the Area 3 RWMS uses subsidence craters formed from underground testing of nuclear weapons for the disposal of packaged and unpackaged bulk waste. The paper describes the technical attributes of both Area 5 and Area 3 facilities, the acceptance process, the disposal processes, and present and future capacities of both sites.

  6. Transuranic contaminated waste form characterization and data base

    SciTech Connect (OSTI)

    McArthur, W.C.; Kniazewycz, B.G.

    1980-07-01

    This report outlines the sources, quantities, characteristics and treatment of transuranic wastes in the United States. This document serves as part of the data base necessary to complete preparation and initiate implementation of transuranic wastes, waste forms, waste container and packaging standards and criteria suitable for inclusion in the present NRC waste management program. No attempt is made to evaluate or analyze the suitability of one technology over another. Indeed, by the nature of this report, there is little critical evaluation or analysis of technologies because such analysis is only appropriate when evaluating a particular application or transuranic waste streams. Due to fiscal restriction, the data base is developed from a myriad of technical sources and does not necessarily contain operating experience and the current status of all technologies. Such an effort was beyond the scope of this report.

  7. Safety analysis report for packaging: the ORNL DOT specification 6M - special form package

    SciTech Connect (OSTI)

    Schaich, R.W.

    1982-07-01

    The ORNL DOT Specification 6M - Special Form Package was fabricated at the Oak Ridge Nation al Laboratory (ORNL) for the transport of Type B solid non-fissile radioactive materials in special form. The package was evaluated on the basis of tests performed by the Dow Chemical Company, Rocky Flats Division, on the DOT-6M container and special form tests performed on a variety of stainless steel capsules at ORNL by Operations Division personnel. The results of these evaluations demonstrate that the package is in compliance with the applicable regulations for the transport of Type B quantities in special form of non-fissile radioactive materials.

  8. Idaho National Engineering Laboratory code assessment of the Rocky Flats transuranic waste

    SciTech Connect (OSTI)

    1995-07-01

    This report is an assessment of the content codes associated with transuranic waste shipped from the Rocky Flats Plant in Golden, Colorado, to INEL. The primary objective of this document is to characterize and describe the transuranic wastes shipped to INEL from Rocky Flats by item description code (IDC). This information will aid INEL in determining if the waste meets the waste acceptance criteria (WAC) of the Waste Isolation Pilot Plant (WIPP). The waste covered by this content code assessment was shipped from Rocky Flats between 1985 and 1989. These years coincide with the dates for information available in the Rocky Flats Solid Waste Information Management System (SWIMS). The majority of waste shipped during this time was certified to the existing WIPP WAC. This waste is referred to as precertified waste. Reassessment of these precertified waste containers is necessary because of changes in the WIPP WAC. To accomplish this assessment, the analytical and process knowledge available on the various IDCs used at Rocky Flats were evaluated. Rocky Flats sources for this information include employee interviews, SWIMS, Transuranic Waste Certification Program, Transuranic Waste Inspection Procedure, Backlog Waste Baseline Books, WIPP Experimental Waste Characterization Program (headspace analysis), and other related documents, procedures, and programs. Summaries are provided of: (a) certification information, (b) waste description, (c) generation source, (d) recovery method, (e) waste packaging and handling information, (f) container preparation information, (g) assay information, (h) inspection information, (i) analytical data, and (j) RCRA characterization.

  9. Robust Solution to Difficult Hydrogen Issues When Shipping Transuranic Waste to the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    Countiss, S. S.; Basabilvazo, G. T.; Moody, D. C. III; Lott, S. A.; Pickerell, M.; Baca, T.; CH2M Hill; Tujague, S.; Svetlik, H.; Hannah, T.

    2003-02-27

    The Waste Isolation Pilot Plant (WIPP) has been open, receiving, and disposing of transuranic (TRU) waste since March 26, 1999. The majority of the waste has a path forward for shipment to and disposal at the WIPP, but there are about two percent (2%) or approximately 3,020 cubic meters (m{sup 3}) of the volume of TRU waste (high wattage TRU waste) that is not shippable because of gas generation limits set by the U.S. Nuclear Regulatory Commission (NRC). This waste includes plutonium-238 waste, solidified organic waste, and other high plutonium-239 wastes. Flammable gases are potentially generated during transport of TRU waste by the radiolysis of hydrogenous materials and therefore, the concentration at the end of the shipping period must be predicted. Two options are currently available to TRU waste sites for solving this problem: (1) gas generation testing on each drum, and (2) waste form modification by repackaging and/or treatment. Repackaging some of the high wattage waste may require up to 20:1 drum increase to meet the gas generation limits of less than five percent (5%) hydrogen in the inner most layer of confinement (the layer closest to the waste). (This is the limit set by the NRC.) These options increase waste handling and transportation risks and there are high costs and potential worker exposure associated with repackaging this high-wattage TRU waste. The U.S. Department of Energy (DOE)'s Carlsbad Field Office (CBFO) is pursuing a twofold approach to develop a shipping path for these wastes. They are: regulatory change and technology development. For the regulatory change, a more detailed knowledge of the high wattage waste (e.g., void volumes, gas generation potential of specific chemical constituents) may allow refinement of the current assumptions in the gas generation model for Safety Analysis Reports for Packaging for Contact-Handled (CH) TRU waste. For technology development, one of the options being pursued is the use of a robust container, the ARROW-PAK{trademark} System. (1) The ARROW-PAK{trademark} is a macroencapsulation treatment technology, developed by Boh Environmental, LLC, New Orleans, Louisiana. This technology has been designed to withstand any unexpected hydrogen deflagration (i.e. no consequence) and other benefits such as criticality control.

  10. Recyclability assessment of nano-reinforced plastic packaging

    SciTech Connect (OSTI)

    SƔnchez, C.; Hortal, M.; Aliaga, C.; Devis, A.; Cloquell-Ballester, V.A.

    2014-12-15

    Highlights: ā€¢ The study compares the recyclability of polymers with and without nanoparticles. ā€¢ Visual appearance, material quality and mechanical properties are evaluated. ā€¢ Minor variations in mechanical properties in R-PE and R-PP with nanoparticles. ā€¢ Slight degradation of R-PET which affect mechanical properties. ā€¢ Colour deviations in recycled PE, PP and PET in ranges higher that 0.3 units. - Abstract: Packaging is expected to become the leading application for nano-composites by 2020 due to the great advantages on mechanical and active properties achieved with these substances. As novel materials, and although there are some current applications in the market, there is still unknown areas under development. One key issue to be addressed is to know more about the implications of the nano-composite packaging materials once they become waste. The present study evaluates the extrusion process of four nanomaterials (Layered silicate modified nanoclay (Nanoclay1), Calcium Carbonate (CaCO{sub 3}), Silver (Ag) and Zinc Oxide (ZnO) as part of different virgin polymer matrices of polyethylene (PE), Polypropylene (PP) and Polyethyleneterephtalate (PET). Thus, the following film plastic materials: (PEā€“Nanoclay1, PEā€“CaCO{sub 3}, PPā€“Ag, PETā€“ZnO, PETā€“Ag, PETā€“Nanoclay1) have been processed considering different recycling scenarios. Results on recyclability show that for PE and PP, in general terms and except for some minor variations in yellowness index, tensile modulus, tensile strength and tear strength (PE with Nanoclay1, PP with Ag), the introduction of nanomaterial in the recycling streams for plastic films does not affect the final recycled plastic material in terms of mechanical properties and material quality compared to conventional recycled plastic. Regarding PET, results show that the increasing addition of nanomaterial into the recycled PET matrix (especially PETā€“Ag) could influence important properties of the recycled material, due to a slight degradation of the polymer, such as increasing pinholes, degradation fumes and elongation at break. Moreover, it should be noted that colour deviations were visible in most of the samples (PE, PP and PET) in levels higher than 0.3 units (limit perceivable by the human eye). The acceptance of these changes in the properties of recycled PE, PP and PET will depend on the specific applications considered (e.g. packaging applications are more strict in material quality that urban furniture or construction products)

  11. HLW Glass Waste Loadings

    Office of Environmental Management (EM)

    HLW Glass Waste Loadings Ian L. Pegg Vitreous State Laboratory The Catholic University of ... (JHCM) technology Factors affecting waste loadings Waste loading requirements ...

  12. Determining site-specific drum loading criteria for storing combustible {sup 238}Pu waste

    SciTech Connect (OSTI)

    Marshall, R.S.; Callis, E.L.; Cappis, J.H.; Espinoza, J.M.; Foltyn, E.M.; Reich, B.T.; Smith, M.C.

    1994-02-01

    Waste containing hydrogenous-combustible material contaminated with {sup 238}Pu can generate hydrogen gas at appreciable rates through alpha radiolysis. To ensure safe transportation of WIPP drums, the limit for {sup 238}Pu-combustible waste published in the WIPP TRUPACT-11 CONTENT (TRUCON) CODES is 21 milliwafts per 55 gallon drum. This corresponds to about 45 milligrams of {sup 238}PuO{sub 2} used for satellite heat source-electrical generators. The Los Alamos waste storage site adopted a {sup 238}Pu waste storage criteria based on these TRCUCON codes. However, reviews of the content in drums of combustible waste generated during heat source assembly at Los Alamos showed the amount of {sup 238}Pu is typically much greater than 45 milligrams. It is not feasible to appreciably reduce Los Alamos {sup 238}Pu waste drum loadings without significantly increasing waste volumes or introducing unsafe practices. To address this concern, a series of studies were implemented to evaluate the applicability of the TRUCON limits for storage of this specific waste. Addressed in these evaluations were determination of the hydrogen generation rate, hydrogen diffusion rates through confinement layers and vent filters, and packaging requirements specific to Los Alamos generated {sup 238}Pu contaminated combustible waste. These studies also showed that the multiple-layer packaging practices in use at Los Alamos could be relaxed without significantly increasing the risk of contamination. Based on a model developed to predict H{sub 2} concentrations in packages and drum headspace, the site specific effective hydrogen generation rate, and hydrogen-diffusion values, and revising the waste packaging practices, we were able to raise the safe loading limit for {sup 238}Pu waste drums for on site storage to the gram levels typical of currently generated {sup 238}Pu waste.

  13. Sampling and analysis plan for Mount Plant D & D soils packages, Revision 1

    SciTech Connect (OSTI)

    1991-02-01

    There are currently 682 containers of soils in storage at Mound Plant, generated between April 1 and October 31, 1990 as a result of excavation of soils containing plutonium-238 at two ongoing Decontamination and Decommissioning (D&D) Program sites. These areas are known as Area 14, the waste transfer system (WTS) hillside, and Area 17, the Special Metallurgical (SM) Building area. The soils from these areas are part of Mound Plant waste stream number AMDM-000000010, Contaminated Soil, and are proposed for shipment to the Nevada Test Site (NTS) for disposal as low-level radioactive waste. The sealed waste packages, constructed of either wood or metal, are currently being stored in Building 31 and at other locations throughout the Mound facility. At a meeting in Las Vegas, Nevada on October, 26, 1990, DOE Nevada Operations Office (DOE-NV) and NTS representatives requested that the Mound Plant D&D soils proposed for shipment to NTS be sampled for Toxicity Characteristic Leaching Procedure (TCLP) constituents. On December 14, 1990, DOE-NV also requested that additional analyses be performed on the soils from one of the soils boxes for polychlorinated biphenyls (PCBs), particle size distribution, and free liquids. The purpose of this plan is to document the proposed sampling and analyses of the packages of D&D soils produced prior to October 31, 1990. In order to provide a thorough description of the soils excavated from the WTS and SM areas, sections 1.1 and 1.2 provide historical Information concerning the D&D soils, including waste stream evaluations and past sampling data.

  14. Step-By-Step Guide for Waste Handling at WIPP - Fact Sheet

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

    the nation's nuclear waste disposal problem Step-By-Step Guide for Waste Handling at WIPP The handling and disposal of contact-handled transuranic waste at the Waste Isolation Pilot Plant (WIPP) involves a series of steps. The following is an overview of those steps. ĀŽ A waste shipment arrives at the WIPP by tractor-trailer. Each tractor-trailer is capable of carrying up to three Transuranic Packaging Transporter Model IIs (TRUPACT-IIs) or HalfPACT's. ĀŽ Upon arrival the tractor trailer and

  15. DEVELOPMENT OF THE BULK TRITIUM SHIPPING PACKAGING

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2008-09-14

    A new radioactive shipping packaging for transporting bulk quantities of tritium, the Bulk Tritium Shipping Package (BTSP), has been designed for the Department of Energy (DOE) as a replacement for a package designed in the early 1970s. This paper summarizes significant design features and describes how the design satisfies the regulatory safety requirements of the Code of Federal Regulations and the International Atomic Energy Agency. The BTSP design incorporates many improvements over its predecessor by implementing improved testing, handling, and maintenance capabilities, while improving manufacturability and incorporating new engineered materials. This paper also discusses the results from testing of the BTSP to 10 CFR 71 Normal Conditions of Transport and Hypothetical Accident Condition events. The programmatic need of the Department of Energy (DOE) to ship bulk quantities of tritium has been satisfied since the late 1970s by the UC-609 shipping package. The current Certificate of Conformance for the UC-609, USA/9932/B(U) (DOE), will expire in late 2011. Since the UC-609 was not designed to meet current regulatory requirements, it will not be recertified and thereby necessitates a replacement Type B shipping package for continued DOE tritium shipments in the future. A replacement tritium packaging called the Bulk Tritium Shipping Package (BTSP) is currently being designed and tested by Savannah River National Laboratory (SRNL). The BTSP consists of two primary assemblies, an outer Drum Assembly and an inner Containment Vessel Assembly (CV), both designed to mitigate damage and to protect the tritium contents from leaking during the regulatory Hypothetical Accident Condition (HAC) events and during Normal Conditions of Transport (NCT). During transport, the CV rests on a silicone pad within the Drum Liner and is covered with a thermal insulating disk within the insulated Drum Assembly. The BTSP packaging weighs approximately 500 lbs without contents and is 50-1/2 inches high by 24-1/2 inches in outside diameter. With contents the gross weight of the BTSP is 650 lbs. The BTSP is designed for the safe shipment of 150 grams of tritium in a solid or gaseous state. To comply with the federal regulations that govern Type B shipping packages, the BTSP is designed so that it will not lose tritium at a rate greater than the limits stated in 10CFR 71.51 of 10{sup -6} A2 per hour for the 'Normal Conditions of Transport' (NCT) and an A2 in 1 week under 'Hypothetical Accident Conditions' (HAC). Additionally, since the BTSP design incorporates a valve as part of the tritium containment boundary, secondary containment features are incorporated in the CV Lid to protect against gas leakage past the valve as required by 10CFR71.43(e). This secondary containment boundary is designed to provide the same level of containment as the primary containment boundary when subjected to the HAC and NCT criteria.

  16. Methods for removing transuranic elements from waste solutions

    SciTech Connect (OSTI)

    Slater, S.A.; Chamberlain, D.B.; Connor, C.; Sedlet, J.; Srinivasan, B.; Vandegrift, G.F.

    1994-11-01

    This report outlines a treatment scheme for separating and concentrating the transuranic (TRU) elements present in aqueous waste solutions stored at Argonne National Laboratory (ANL). The treatment method selected is carrier precipitation. Potential carriers will be evaluated in future laboratory work, beginning with ferric hydroxide and magnetite. The process will result in a supernatant with alpha activity low enough that it can be treated in the existing evaporator/concentrator at ANL. The separated TRU waste will be packaged for shipment to the Waste Isolation Pilot Plant.

  17. Hawaii Information Package for Chemical Inventory Form (HCIF...

    Open Energy Info (EERE)

    Information Package for Chemical Inventory Form (HCIF)Tier II Jump to: navigation, search OpenEI Reference LibraryAdd to library Form: Hawaii Information Package for Chemical...

  18. Bi-level microelectronic device package with an integral window

    DOE Patents [OSTI]

    Peterson, Kenneth A.; Watson, Robert D.

    2004-01-06

    A package with an integral window for housing a microelectronic device. The integral window is bonded directly to the package without having a separate layer of adhesive material disposed in-between the window and the package. The device can be a semiconductor chip, CCD chip, CMOS chip, VCSEL chip, laser diode, MEMS device, or IMEMS device. The multilayered package can be formed of a LTCC or HTCC cofired ceramic material, with the integral window being simultaneously joined to the package during LTCC or HTCC processing. The microelectronic device can be flip-chip bonded so that the light-sensitive side is optically accessible through the window. The package has at least two levels of circuits for making electrical interconnections to a pair of microelectronic devices. The result is a compact, low-profile package having an integral window that is hermetically sealed to the package prior to mounting and interconnecting the microelectronic device(s).

  19. [PATCHED] python/2.7.4 gzip package fails

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

    PATCHED python2.7.4 gzip package fails PATCHED python2.7.4 gzip package fails September 24, 2013 by Doug Jacobsen The modules version of python (python2.7.4) had a bug in...

  20. Waste Hoist

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

    Primary Hoist: 45-ton Rope-Guide Friction Hoist Largest friction hoist in the world when it was built in 1985 Completely enclosed (for contamination control), the waste hoist at WIPP is a modern friction hoist with rope guides (uses a balanced counterweight and tail ropes). With a 45-ton capacity, it was the largest friction hoist in the world when it was built in 1986. Hoist deck footprint: 2.87m wide x 4.67m long Hoist deck height: 2.87m wide x 7.46m high Access height to the waste hoist deck

  1. CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

    SciTech Connect (OSTI)

    Brinkman, K.; Fox, K.; Marra, J.

    2012-05-15

    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.

  2. RECENT PROGRESS IN DOE WASTE TANK CLOSURE

    SciTech Connect (OSTI)

    Langton, C

    2008-02-01

    The USDOE complex currently has over 330 underground storage tanks that have been used to process and store radioactive waste generated from the production of weapons materials. These tanks contain over 380 million liters of high-level and low-level radioactive waste. The waste consists of radioactively contaminated sludge, supernate, salt cake or calcine. Most of the waste exists at four USDOE locations, the Hanford Site, the Savannah River Site, the Idaho Nuclear Technology and Engineering Center and the West Valley Demonstration Project. A summary of the DOE tank closure activities was first issued in 2001. Since then, regulatory changes have taken place that affect some of the sites and considerable progress has been made in closing tanks. This paper presents an overview of the current regulatory changes and drivers and a summary of the progress in tank closures at the various sites over the intervening six years. A number of areas are addressed including closure strategies, characterization of bulk waste and residual heel material, waste removal technologies for bulk waste, heel residuals and annuli, tank fill materials, closure system modeling and performance assessment programs, lessons learned, and external reviews.

  3. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM - 2011

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2012-06-21

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2011 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2011 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per SRR-LWE-2011-00026, HLW Tank Farm Inspection Plan for 2011, were completed. Ultrasonic measurements (UT) performed in 2011 met the requirements of C-ESR-G-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 3, and WSRC-TR-2002-00061, Rev.6. UT inspections were performed on Tanks 25, 26 and 34 and the findings are documented in SRNL-STI-2011-00495, Tank Inspection NDE Results for Fiscal Year 2011, Waste Tanks 25, 26, 34 and 41. A total of 5813 photographs were made and 835 visual and video inspections were performed during 2011. A potential leaksite was discovered at Tank 4 during routine annual inspections performed in 2011. The new crack, which is above the allowable fill level, resulted in no release to the environment or tank annulus. The location of the crack is documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.6.

  4. Waste processing air cleaning

    SciTech Connect (OSTI)

    Kriskovich, J.R.

    1998-07-27

    Waste processing and preparing waste to support waste processing relies heavily on ventilation. Ventilation is used at the Hanford Site on the waste storage tanks to provide confinement, cooling, and removal of flammable gases.

  5. LAC Regional Platform Workshop Participant Package | Open Energy...

    Open Energy Info (EERE)

    Announcement Agenda Participant Package Accommodations Location & Transportation Insurance & Visas Participants Presentations Outcomes Report Links Contact Us General...

  6. Enforcement Letter, Packaging Specialties, Inc. - April 6, 2015 |

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

    Department of Energy Packaging Specialties, Inc. - April 6, 2015 Enforcement Letter, Packaging Specialties, Inc. - April 6, 2015 April 6, 2015 Nuclear Safety Enforcement Letter issued to Packaging Specialties, Inc., regarding deficiencies in testing, inspection, and certification of shipping containers used at DOE's Pantex Plant. On April 6, 2015, the U.S. Department of Energy's (DOE) Office of Enforcement issued an Enforcement Letter (NEL-2015-02) to Packaging Specialties, Inc. (PSI) for

  7. Human Portable Radiation Detection System Communications Package Evaluation

    SciTech Connect (OSTI)

    Morgen, Gerald P.; Peterson, William W.

    2009-06-11

    Testing and valuation of the Human Portable Radiation Detection System Communications Package for the US Coast Guard.

  8. Performance of Conduction Cooled Splittable Superconducting Magnet Package

    Office of Scientific and Technical Information (OSTI)

    for Linear Accelerators (Conference) | SciTech Connect Performance of Conduction Cooled Splittable Superconducting Magnet Package for Linear Accelerators Citation Details In-Document Search Title: Performance of Conduction Cooled Splittable Superconducting Magnet Package for Linear Accelerators New Linear Superconducting Accelerators need a superconducting magnet package installed inside SCRF Cryomodules to focus and steer electron or proton beams. A superconducting magnet package was

  9. Safety analysis report for packaging (onsite) steel drum

    SciTech Connect (OSTI)

    McCormick, W.A.

    1998-09-29

    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.

  10. CD Label and Package Templates | Department of Energy

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

    Publications, Exhibits, & Logos Ā» Templates Ā» CD Label and Package Templates CD Label and Package Templates The Office of Energy Efficiency and Renewable Energy (EERE) has developed templates for CD labels and CD packages. These can be used for all EERE products. Both templates are available as EPS files, which can be downloaded and edited in a graphics package like Adobe Illustrator. You are not required to use these templates for your EERE products. These templates were designed to allow

  11. CHP Integrated with Packaged Boilers - Presentation by CMCE, Inc., June

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

    2011 | Department of Energy Integrated with Packaged Boilers - Presentation by CMCE, Inc., June 2011 CHP Integrated with Packaged Boilers - Presentation by CMCE, Inc., June 2011 Presentation on CHP Integrated with Packaged Boilers, given by Carlo Castaldini of CMCE, Inc., at the U.S. DOE Industrial Distributed Energy Portfolio Review Meeting in Washington, D.C. on June 1-2, 2011. PDF icon packaged_boilers_castaldini.pdf More Documents & Publications CHP Integrated with Burners for

  12. Environmental restoration risk-based prioritization work package planning and risk ranking methodology. Revision 2

    SciTech Connect (OSTI)

    Dail, J.L.; Nanstad, L.D.; White, R.K.

    1995-06-01

    This document presents the risk-based prioritization methodology developed to evaluate and rank Environmental Restoration (ER) work packages at the five US Department of Energy, Oak Ridge Field Office (DOE-ORO) sites [i.e., Oak Ridge K-25 Site (K-25), Portsmouth Gaseous Diffusion Plant (PORTS), Paducah Gaseous Diffusion Plant (PGDP), Oak Ridge National Laboratory (ORNL), and the Oak Ridge Y-12 Plant (Y-12)], the ER Off-site Program, and Central ER. This prioritization methodology was developed to support the increased rigor and formality of work planning in the overall conduct of operations within the DOE-ORO ER Program. Prioritization is conducted as an integral component of the fiscal ER funding cycle to establish program budget priorities. The purpose of the ER risk-based prioritization methodology is to provide ER management with the tools and processes needed to evaluate, compare, prioritize, and justify fiscal budget decisions for a diverse set of remedial action, decontamination and decommissioning, and waste management activities. The methodology provides the ER Program with a framework for (1) organizing information about identified DOE-ORO environmental problems, (2) generating qualitative assessments of the long- and short-term risks posed by DOE-ORO environmental problems, and (3) evaluating the benefits associated with candidate work packages designed to reduce those risks. Prioritization is conducted to rank ER work packages on the basis of the overall value (e.g., risk reduction, stakeholder confidence) each package provides to the ER Program. Application of the methodology yields individual work package ``scores`` and rankings that are used to develop fiscal budget requests. This document presents the technical basis for the decision support tools and process.

  13. Basalt Waste Isolation Project. Annual report, fiscal year 1980

    SciTech Connect (OSTI)

    Not Available

    1980-11-01

    During this fiscal year the information available in the fields of geology and hydrology of the Columbia Plateau was consolidated and two reports were issued summarizing this information. In addition, the information on engineered barriers was consolidated and a report summarizing the research to date on waste package development and design of borehole seals was prepared. The waste package studies, when combined with the hydrologic integration, revealed that even under extreme disruptive conditions, a repository in basalt with appropriately designed waste packages can serve as an excellent barrier for containment of radionuclides for the long periods of time required for waste isolation. On July 1, 1980, the first two heater tests at the Near-Surface Test Facility were started and have been successfully operated to this date. The papers on the Near-Surface Test Facility section of this report present the results of the equipment installed and the preliminary results of the testing. In October 1979, the US Department of Energy selected the joint venture of Kaiser Engineers/Parsons Brinckerhoff Quade and Douglas, Inc., to be the architect-engineer to produce a conceptual design of a repository in basalt. During the year, this design has progressed and concept selection has now been completed. This annual report presents a summary of the highlights of the work completed during fiscal year 1980. It is intended to supplement and summarize the nearly 200 papers and reports that have been distributed to date as a part of the Basalt Waste Isolation Project studies.

  14. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

    SciTech Connect (OSTI)

    1994-10-01

    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.

  15. Lattice thermal conductivity of filled skutterudites: An anharmonicity perspective

    SciTech Connect (OSTI)

    Geng, Huiyuan, E-mail: genghuiyuan@hit.edu.cn; Meng, Xianfu; Zhang, Hao; Zhang, Jian [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China)

    2014-10-28

    We report a phenomenological model to calculate the high-temperature lattice thermal conductivity of filled skutterudite antimonides. The model needs no phonon resonant scattering terms. Instead, we assume that umklapp processes dominate the high-temperature phonon scattering. In order to represent the anharmonicity introduced by the filling atom, we introduce a Gaussian term into the relaxation time of the umklapp process. The developed model agrees remarkably well with the experimental results of RE{sub f}Co{sub 4}Sb{sub 12} and RE{sub f}Fe{sub 4}Sb{sub 12} (RE?=?Yb, Ba, and Ca) alloys. To further test the validity of our model, we calculate the lattice thermal conductivity of nanostructured or multi-filled skutterudites. The calculation results are also in good agreement with experiment, increasing our confidence in the developed anharmonicity model.

  16. ALTERNATE MATERIALS IN DESIGN OF RADIOACTIVE MATERIAL PACKAGES

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2010-07-09

    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.

  17. Inventory Data Package for Hanford Assessments

    SciTech Connect (OSTI)

    Kincaid, Charles T.; Eslinger, Paul W.; Aaberg, Rosanne L.; Miley, Terri B.; Nelson, Iral C.; Strenge, Dennis L.; Evans, John C.

    2006-06-01

    This document presents the basis for a compilation of inventory for radioactive contaminants of interest by year for all potentially impactive waste sites on the Hanford Site for which inventory data exist in records or could be reasonably estimated. This document also includes discussions of the historical, current, and reasonably foreseeable (1944 to 2070) future radioactive waste and waste sites; the inventories of radionuclides that may have a potential for environmental impacts; a description of the method(s) for estimating inventories where records are inadequate; a description of the screening method(s) used to select those sites and contaminants that might make a substantial contribution to impacts; a listing of the remedial actions and their completion dates for waste sites; and tables showing the best estimate inventories available for Hanford assessments.

  18. Fractional Quantum Hall Effect at Landau Level Filling v=4/11...

    Office of Scientific and Technical Information (OSTI)

    Fractional Quantum Hall Effect at Landau Level Filling v411. Citation Details In-Document Search Title: Fractional Quantum Hall Effect at Landau Level Filling v411. Abstract...

  19. Waste drum refurbishment

    SciTech Connect (OSTI)

    Whitmill, L.J.

    1996-10-18

    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.

  20. Packaging design criteria for the MCO cask

    SciTech Connect (OSTI)

    Edwards, W.S.

    1996-04-29

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins (including possibly 700 additional elements from PUREX, N Reactor, and 327 Laboratory). The basin water, particularly in the K East Basin, contains significant quantities of dissolved nuclear isotopes and radioactive fuel corrosion particles. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) in the 200 East area. In order to initiate K Basin cleanup on schedule, the two-year fuel-shipping campaign must begin by December 1997. The purpose of this packaging design criteria is to provide criteria for the design, fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multiple Canister Overpacks.