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Sample records for canister overpack mco

  1. Multi Canister Overpack (MCO) Design Report [SEC 1 Thru 3

    Energy Technology Data Exchange (ETDEWEB)

    GOLDMANN, L.H.

    2000-02-29

    The MCO is designed to facilitate the removal, processing and storage of the spent nuclear fuel currently stored in the East and West K-Basins. The MCO is a stainless steel canister approximately 24 inches in diameter and 166 inches long with cover cap installed. The shell and the collar which is welded to the shell are fabricated from 304/304L dual certified stainless steel for the shell and F304/F304L dual certified for the collar. The shell has a nominal thickness of 1/2 inch. The top closure consists of a shield plug with four processing ports and a locking ring with jacking bolts to pre-load a metal seal under the shield plug. The fuel is placed in one of four types of baskets, excluding the SPR fuel baskets, in the fuel retention basin. Each basket is then loaded into the MCO which is inside the transfer cask. Once all of the baskets are loaded into the MCO, the shield plug with a process tube is placed into the open end of the MCO. This shield plug provides shielding for workers when the transfer cask, containing the MCO, is lifted from the pool. After being removed from the pool, the locking ring is installed and the jacking bolts are tightened to pre-load the metal main closure seal. The cask is then sealed and the MCO taken to the Cold Vacuum Drying (CVD) facility for bulk water removal and vacuum drying through the process ports. Covers for the process ports may be installed or removed as needed per operating procedures. The MCO is then transferred to the Canister Storage Building (CSB), in the closed transfer cask. At the CSB, the MCO is then removed from the cask and becomes one of two MCOs stacked in a storage tube. MCOs will have a cover cap welded over the shield plug providing a complete welded closure. A number of MCOs may be stored with just the mechanical seal to allow monitoring of the MCO pressure, temperature, and gas composition.

  2. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

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    HOLLENBECK, R.G.

    2000-05-08

    The Spent Nuclear Fuel (SNF) Canister Storage Building (CSB) is the interim storage facility for the K-Basin SNF at the US. Department of Energy (DOE) Hanford Site. The SNF is packaged in multi-canister overpacks (MCOs). The MCOs are placed inside transport casks, then delivered to the service station inside the CSB. At the service station, the MCO handling machine (MHM) moves the MCO from the cask to a storage tube or one of two sample/weld stations. There are 220 standard storage tubes and six overpack storage tubes in a below grade reinforced concrete vault. Each storage tube can hold two MCOs.

  3. Multi Canister Overpack (MCO) Topical Report [SEC 1 THRU 3

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    LORENZ, B.D.

    2000-05-11

    In February 1995, the US Department of Energy (DOE) approved the Spent Nuclear Fuel (SNF) Project's ''Path Forward'' recommendation for resolution of the safety and environmental concerns associated with the deteriorating SNF stored in the Hanford Site's K Basins (Hansen 1995). The recommendation included an aggressive series of projects to design, construct, and operate systems and facilitates to permit the safe retrieval, packaging, transport, conditions, and interim storage of the K Basins' SNF. The facilities are the Cold VAcuum Drying Facility (CVDF) in the 100 K Area of the Hanford Site and the Canister Storage building (CSB) in the 200 East Area. The K Basins' SNF is to be cleaned, repackaged in multi-canister overpacks (MCOs), removed from the K Basins, and transported to the CVDF for initial drying. The MCOs would then be moved to the CSB and weld sealed (Loscoe 1996) for interim storage (about 40 years). One of the major tasks associated with the initial Path Forward activities is the development and maintenance of the safety documentation. In addition to meeting the construction needs for new structures, the safety documentation for each must be generated.

  4. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

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    HOLLENBECK, R.G.

    2000-06-07

    The purpose of this report is to investigate the potential for damage to the multi-canister overpack (MCO) during impact from an eccentric accidental drop onto the standard storage tube, overpack storage tube, service station or sampling/weld station. Damage to the storage tube and sample/weld station is beyond the scope of this report. The results of this analysis are required to show the following: (1) If a breach resulting in unacceptable release of contamination could occur in the MCO. (2) If the dropped MCO could become stuck in the storage tube after the drop. (3) Maximum deceleration of the spent nuclear fuel baskets. The model appropriate for the standard storage tubes with the smaller gap is the basis for the analysis and results reported herein in this SNF-5204, revision 2 report. Revision 1 of this report is based on a model that includes the larger gap appropriate for the overpack tubes.

  5. Estimates of Particulate Mass in Multi Canister Overpacks (MCO)

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    SLOUGHTER, J.P.

    2000-02-16

    High, best estimate, and low values are developed for particulate inventories within MCO baskets that have been loaded with freshly cleaned fuel assemblies and scrap. These per-basket estimates are then applied to all anticipated MCO payload configurations to identify which configurations are bounding for each type of particulate. Finally the resulting bounding and nominal values for residual particulates are combined with corresponding values [from other documents] for particulates that may be generated by corrosion of exposed uranium after the fuel has been cleaned. The resulting rounded nominal estimate for a typical MCO after 40 years of storage is 8 kg. The estimate for a bounding total particulate case MCO is that it may contain up to 64 kg of particulate after 40 years of storage.

  6. Multi Canister Overpack (MCO) Combustible Gas Management Leak Test Acceptance Criteria (OCRWM)

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    SHERRELL, D.L.

    2000-10-10

    The purpose of this document is to support the Spent Nuclear Fuel Project's combustible gas management strategy while avoiding the need to impose any requirements for oxygen free atmospheres within storage tubes that contain multi-canister overpacks (MCO). In order to avoid inerting requirements it is necessary to establish and confirm leak test acceptance criteria for mechanically sealed and weld sealed MCOs that are adequte to ensure that, in the unlikely event the leak test results for any MCO were to approach either of those criteria, it could still be handled and stored in stagnant air without compromising the SNF Project's overall strategy to prevent accumulation of combustible gas mixtures within MCOs or within their surroundings. To support that strategy, this document: (1) establishes combustible gas management functions and minimum functional requirements for the MCO's mechanical seals and closure weld(s); (2) establishes a maximum practical value for the minimum required initial MCO inert backfill gas pressure; and (3) based on items 1 and 2, establishes and confirms leak test acceptance criteria for the MCO's mechanical seal and final closure weld(s).

  7. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

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    TU, K.C.

    1999-10-08

    Multi-Canister Overpacks (MCOs) containing spent nuclear fuel (SNF) will be routinely handled at the Canister Storage Building (CSB) during fuel movement operations in the SNF Project. This analysis was performed to investigate the potential for damage from an eccentric accidental drop onto the standard storage tube, overpack tube, service station, or sample/weld station. Appendix D was added to the FDNW document to include the peer Review Comment Record & transmittal record.

  8. Multi Canister Overpack (MCO) Handling Machine Independent Review of Seismic Structural Analysis

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    SWENSON, C.E.

    2000-09-22

    The following separate reports and correspondence pertains to the independent review of the seismic analysis. The original analysis was performed by GEC-Alsthom Engineering Systems Limited (GEC-ESL) under subcontract to Foster-Wheeler Environmental Corporation (FWEC) who was the prime integration contractor to the Spent Nuclear Fuel Project for the Multi-Canister Overpack (MCO) Handling Machine (MHM). The original analysis was performed to the Design Basis Earthquake (DBE) response spectra using 5% damping as required in specification, HNF-S-0468 for the 90% Design Report in June 1997. The independent review was performed by Fluor-Daniel (Irvine) under a separate task from their scope as Architect-Engineer of the Canister Storage Building (CSB) in 1997. The comments were issued in April 1998. Later in 1997, the response spectra of the Canister Storage Building (CSB) was revised according to a new soil-structure interaction analysis and accordingly revised the response spectra for the MHM and utilized 7% damping in accordance with American Society of Mechanical Engineers (ASME) NOG-1, ''Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).'' The analysis was re-performed to check critical areas but because manufacturing was underway, designs were not altered unless necessary. FWEC responded to SNF Project correspondence on the review comments in two separate letters enclosed. The dispositions were reviewed and accepted. Attached are supplier source surveillance reports on the procedures and process by the engineering group performing the analysis and structural design. All calculation and analysis results are contained in the MHM Final Design Report which is part of the Vendor Information File 50100. Subsequent to the MHM supplier engineering analysis, there was a separate analyses for nuclear safety accident concerns that used the electronic input data files provided by FWEC/GEC-ESL and are contained in

  9. FEMA and RAM Analysis for the Multi Canister Overpack (MCO) Handling Machine

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    SWENSON, C.E.

    2000-06-01

    The Failure Modes and Effects Analysis and the Reliability, Availability, and Maintainability Analysis performed for the Multi-Canister Overpack Handling Machine (MHM) has shown that the current design provides for a safe system, but the reliability of the system (primarily due to the complexity of the interlocks and permissive controls) is relatively low. No specific failure modes were identified where significant consequences to the public occurred, or where significant impact to nearby workers should be expected. The overall reliability calculation for the MHM shows a 98.1 percent probability of operating for eight hours without failure, and an availability of the MHM of 90 percent. The majority of the reliability issues are found in the interlocks and controls. The availability of appropriate spare parts and maintenance personnel, coupled with well written operating procedures, will play a more important role in successful mission completion for the MHM than other less complicated systems.

  10. Criticality Safety Evaluation Report CSER-96-019 for Spent Nuclear Fuel (SNF) Processing and Storage Facilities Multi Canister Overpack (MCO)

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    KESSLER, S.F.

    1999-10-20

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operation at the Cold Vacuum Drying Facility,a nd storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the K{sub eff} = 0.95 criticality safety limit. This revision incorporates the analyses for the sampling/weld station in the Canister Storage Building and additional analysis of the MCO during the draining at CVDF. Additional discussion of the scrap basket model was added to show why the addition of copper divider plates was not included in the models.

  11. Evaluation of Multi Canister Overpack (MCO) Handling Machine Uplift Restraint for a Seismic Event During Repositioning Operations

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    SWENSON, C.E.

    2000-05-15

    Insertion of the Multi-Canister Overpack (MCO) assemblies into the Canister Storage Building (CSB) storage tubes involves the use of the MCO Handling Machine (MHM). During MCO storage tube insertion operations, inadvertent movement of the MHM is prevented by engaging seismic restraints (''active restraints'') located adjacent to both the bridge and trolley wheels. During MHM repositioning operations, the active restraints are not engaged. When the active seismic restraints are not engaged, the only functioning seismic restraints are non-engageable (''passive'') wheel uplift restraints which function only if the wheel uplift is sufficient to close the nominal 0.5-inch gap at the uplift restraint interface. The MHM was designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis reported seismic loads on the MHM uplift restraints and EDERER performed corresponding structural calculations to demonstrate structural adequacy of the seismic uplift restraint hardware. The ALSTHOM and EDERER calculations were performed for a parked MHM with the active seismic restraints engaged, resulting in uplift restraint loading only in the vertical direction. In support of development of the CSB Safety Analysis Report (SAR), an evaluation of the MHM seismic response was requested for the case where the active seismic restraints are not engaged. If a seismic event occurs during MHM repositioning operations, a moving contact at a seismic uplift restraint would introduce a friction load on the restraint in the direction of the movement. These potential horizontal friction loads on the uplift restraints were not included in the existing restraint hardware design calculations. One of the purposes of the current evaluation is to address the structural adequacy of the MHM seismic uplift restraints with the addition of the horizontal friction associated with MHM repositioning movements.

  12. Impact of Aluminum on Anticipated Corrosion in a Flooded SNF Multi Canister Overpack (MCO)

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    DUNCAN, D.R.

    1999-07-06

    Corrosion reactions in a flooded MCO are examined to determine the impact of aluminum corrosion products (from aluminum basket grids and spacers) on bound water estimates and subsequent fuel/environment reactions during storage. The mass and impact of corrosion products were determined to be insignificant, validating the choice of aluminum as an MCO component and confirming expectations that no changes to the Technical Databook or particulate mass or water content are necessary.

  13. Drop Testing Representative Multi-Canister Overpacks

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    Snow, Spencer D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morton, Dana K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    The objective of the work reported herein was to determine the ability of the Multi- Canister Overpack (MCO) canister design to maintain its containment boundary after an accidental drop event. Two test MCO canisters were assembled at Hanford, prepared for testing at the Idaho National Engineering and Environmental Laboratory (INEEL), drop tested at Sandia National Laboratories, and evaluated back at the INEEL. In addition to the actual testing efforts, finite element plastic analysis techniques were used to make both pre-test and post-test predictions of the test MCOs structural deformations. The completed effort has demonstrated that the canister design is capable of maintaining a 50 psig pressure boundary after drop testing. Based on helium leak testing methods, one test MCO was determined to have a leakage rate not greater than 1x10-5 std cc/sec (prior internal helium presence prevented a more rigorous test) and the remaining test MCO had a measured leakage rate less than 1x10-7 std cc/sec (i.e., a leaktight containment) after the drop test. The effort has also demonstrated the capability of finite element methods using plastic analysis techniques to accurately predict the structural deformations of canisters subjected to an accidental drop event.

  14. Criticality safety evaluation report for the multi-canister overpack

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    KESSLER, S.F.

    1999-05-21

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark 1V or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operations at the Cold Vacuum Drying Facility, and storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} = 0.95 criticality safety limit. Additional analyses in this revision include partial fuel basket loadings, loading 26.1 inch Mark IA fuel assemblies into Mark IV fuel baskets, and the revised fuel and scrap basket designs. The MCO MCNP model was revised to include the shield plug assembly.

  15. Multi-canister overpack design report

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    SMITH, K.E.

    1999-05-12

    Revision 2 incorporates changes to reflect a 150 psig pressure rating for the mechanically closed MCO and 450 psig pressure rating with the cover cap welded in place, per the MCO Performance Specification, HNF-S-0426, Rev. 5 .

  16. Estimates of particulate mass in multi-canister overpacks

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    SLOUGHTER, J.P.

    1999-02-25

    High, best estimate, and low values are developed for particulate inventories within MCO baskets that have been loaded with freshly cleaned fuel assemblies and scrap. These per-basket estimates are then applied to all anticipated MCO payload configurations to identify which configurations are bounding for each type of particulate. Finally the resulting bounding and nominal values for residual particulates are combined with corresponding values [from other documents] for particulate that may be generated by corrosion of exposed uranium after the fuel has been cleaned. The resulting rounded nominal estimate for a typical MCO after 40 years of storage is 8 kg. The estimate for a bounding total particulate case MCO is that it may contain up to 64 kg of particulate after 40 years of storage.

  17. ASME Code requirements for multi-canister overpack design and fabrication

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    SMITH, K.E.

    1998-11-03

    The baseline requirements for the design and fabrication of the MCO include the application of the technical requirements of the ASME Code, Section III, Subsection NB for containment and Section III, Subsection NG for criticality control. ASME Code administrative requirements, which have not historically been applied at the Hanford site and which have not been required by the US Nuclear Regulatory Commission (NRC) for licensed spent fuel casks/canisters, were not invoked for the MCO. As a result of recommendations made from an ASME Code consultant in response to DNFSB staff concerns regarding ASME Code application, the SNF Project will be making the following modifications: issue an ASME Code Design Specification and Design Report, certified by a Registered Professional Engineer; Require the MCO fabricator to hold ASME Section III or Section VIII, Division 2 accreditation; and Use ASME Authorized Inspectors for MCO fabrication. Incorporation of these modifications will ensure that the MCO is designed and fabricated in accordance with the ASME Code. Code Stamping has not been a requirement at the Hanford site, nor for NRC licensed spent fuel casks/canisters, but will be considered if determined to be economically justified.

  18. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

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    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  19. MCO loading and cask loadout technical manual

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    PRAGA, A.N.

    1998-10-01

    A compilation of the technical basis for loading a multi-canister overpack (MCO) with spent nuclear fuel and then placing the MCO into a cask for shipment to the Cold Vacuum Drying Facility. The technical basis includes a description of the process, process technology that forms the basis for loading alternatives, process control considerations, safety considerations, equipment description, and a brief facility structure description.

  20. Spent Nuclear Fuel Project (SNFP) gas generation from N-Fuel in multi-canister overpacks

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    Cooper, T.D.

    1996-08-01

    During the conversion from wet pool storage for spent nuclear fuel at Hanford, gases will be generated from both radiolysis and chemical reactions. The gas generation phenomenon needs to be understood as it applies to safety and design issues,specifically over pressurization of sealed storage containers,and detonation/deflagration of flammable gases. This study provides an initial basis to predict the implications of gas generation on the proposed functional processes for spent nuclear fuel conversion from wet to dry storage. These projections are based upon examination of the history of fuel manufacture at Hanford, irradiation in the reactors, corrosion during wet pool storage, available fuel characterization data and available information from literature. Gas generation via radiolysis and metal corrosion are addressed. The study examines gas generation, the boundary conditions for low medium and high levels of sludge in SNF storage/processing containers. The functional areas examined include: flooded and drained Multi-Canister Overpacks, cold vacuum drying, shipping and staging and long term storage.

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

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    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. DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

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    R.J. Garrett

    2001-07-30

    The purpose of this calculation is to provide the bases for defining the preclosure limits on radioactive material releases from radioactive waste forms to be received in disposable canisters at the Monitored Geologic Repository (MGR) at Yucca Mountain. Specifically, this calculation will provide the basis for criteria to be included in a forthcoming revision of the Waste Acceptance System Requirements Document (WASRD) that limits releases in terms of non-isotope-specific canister release dose-equivalent source terms. These criteria will be developed for the Department of Energy spent nuclear fuel (DSNF) standard canister, the Multicanister Overpack (MCO), the naval spent fuel canister, the High-Level Waste (HLW) canister, the plutonium can-in-canister, and the large Multipurpose Canister (MPC). The shippers of such canisters will be required to demonstrate that they meet these criteria before the canisters are accepted at the MGR. The Quality Assurance program is applicable to this calculation. The work reported in this document is part of the analysis of DSNF and is performed using procedure AP-3.124, Calculations. The work done for this analysis was evaluated according to procedure QAP-2-0, Control of Activities, which has been superseded by AP-2.21Q, Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities. This evaluation determined that such activities are subject to the requirements of DOE/RW/0333P, Quality Assurance Requirements and Description (DOE 2000). This work is also prepared in accordance with the development plan titled Design Basis Event Analyses on DOE SNF and Plutonium Can-In-Canister Waste Forms (CRWMS M&O 1999a) and Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages (CRWMS M&O 2000d). This calculation contains no electronic data applicable to any electronic data management system.

  3. MCO combustible gas management leak test acceptance criteria

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    SHERRELL, D.L.

    1999-05-11

    Existing leak test acceptance criteria for mechanically sealed and weld sealed multi-canister overpacks (MCO) were evaluated to ensure that MCOs can be handled and stored in stagnant air without compromising the Spent Nuclear Fuel Project's overall strategy to prevent accumulation of combustible gas mixtures within MCO's or within their surroundings. The document concludes that the integrated leak test acceptance criteria for mechanically sealed and weld sealed MCOs (1 x 10{sup -5} std cc/sec and 1 x 10{sup -7} std cc/sec, respectively) are adequate to meet all current and foreseeable needs of the project, including capability to demonstrate compliance with the NFPA 60 Paragraph 3-3 requirement to maintain hydrogen concentrations [within the air atmosphere CSB tubes] t or below 1 vol% (i.e., at or below 25% of the LFL).

  4. Statement of work for sytem design and engineering of the spent nuclear fuel multi-cansiter overpack

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    Smith, K.E., Fluor Daniel Hanford

    1997-03-03

    This Statement of Work (SOW) describes the work scope for the preparation of the Phase 2 (final) design for the Multiple Canister Overpack (MCO) equipment. The MCO is to be used as the radiological containment device for the Spent Nuclear Fuel (SNF) assemblies, currently in wet storage in K East and West Basins, to be transported and stored in the Canister Storage Building (CSB) until final disposal facilities are made available. The engineering services contractor will be requested to provide reports, studies, analyses, engineering, drawings, specifications, estimates and schedules. The overall goal of this task order is to do the following: 1. Prepare a fabrication specification, ASME Code exception report, a packaging, shipping and warehouse plan, and detailed fabrication drawings of the MCO in accordance with the MCO Performance Specification (HNF-S-0426, Rev. 3) for procurement activities by the SNF MCO Subproject. 2. Establish and maintain a comment data base on the comments, resolutions, changes to the design of the MCO. 3. Support fabrication activities through the review of vendor fabrication drawings and shop test reports.

  5. Evaluation of copper for divider subassembly in MCO Mark IA and Mark IV scrap fuel baskets

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    Graves, C.E.

    1997-09-29

    The K Basin Spent Nuclear Fuel (SNF) Project Multi-Canister Overpack (MCO) subprojection eludes the design and fabrication of a canister that will be used to confine, contain, and maintain fuel in a critically safe array to enable its removal from the K Basins, vacuum drying, transport, staging, hot conditioning, and interim storage (Goldinann 1997). Each MCO consists of a shell, shield plug, fuel baskets (Mark IA or Mark IV), and other incidental equipment. The Mark IA intact and scrap fuel baskets are a safety class item for criticality control and components necessary for criticality control will be constructed from 304L stainless steel. It is proposed that a copper divider subassembly be used in both Mark IA and Mark IV scrap baskets to increase the safety basis margin during cold vacuum drying. The use of copper would increase the heat conducted away from hot areas in the baskets out to the wall of the MCO by both radiative and conductive heat transfer means. Thus copper subassembly will likely be a safety significant component of the scrap fuel baskets. This report examines the structural, cost and corrosion consequences associated with using a copper subassembly in the stainless steel MCO scrap fuel baskets.

  6. Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)

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    POWERS, T.B.

    2000-03-20

    By using simple frequency calculations and fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The following are the design basis accidents: Mechanical damage of MCO; Gaseous release from the MCO; MCO internal hydrogen deflagration; MCO external hydrogen deflagration; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

  7. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

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    KLEM, M.J.

    2000-10-18

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8

  8. Shippingport Spent Fuel Canister (SSFC) Design Report Project W-518

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    JOHNSON, D.M.

    2000-01-27

    The SSFC Design Report Describes A spent fuel canister for Shippingport Core 2 blanket fuel assemblies. The design of the SSFC is a minor modification of the MCO. The modification is limited to the Shield Plug which remains unchanged with regard to interfaces with the canister shell. The performance characteristics remain those for the MCO, which bounds the payload of the SSFC.

  9. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

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    PICKETT, W.W.

    2000-09-22

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

  10. Packaging design criteria for the MCO cask

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    Edwards, W.S.

    1997-01-30

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the K Basins to a Canister Storage Building in the 200 East Area. 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.

  11. Multicanister overpack topical report

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    Lorenz, B.D., Fluor Daniel Hanford

    1997-03-25

    The Spent Nuclear Fuel MCO is a single-use container that consists of a cylindrical shell, five to six fuel baskets, a shield plug, and features necessary for maintaining the structural integrity of the MCO while providing criticality control and fuel processing capability.

  12. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those

  13. Assessment study of the stresses induced by corrosion in the Advanced Cold Process Canister

    Energy Technology Data Exchange (ETDEWEB)

    Hoch, A.R.; Sharland, S.M. [Chemical Studies Department, Radwaste Disposal Division, AEA Decommissioning and Radwaste, Harwell Laboratory, Oxfordshire (United Kingdom)

    1993-10-01

    The Advanced Cold Process Canister (ACPC) is a concept for the encapsulation of spent nuclear fuel for geological disposal. The basic design of the ACPC consists of an outer oxygen free copper overpack covering a carbon steel inner container. In this report the stresses exerted on the copper overpack as a result of an early failure of the canister and the subsequent corrosion of the steel are calculated. 4 figs, 8 refs, 2 tabs.

  14. Evaluation of Impact Resistance of Concrete Overpack of Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sanghoon; Kim, Ki-Young; Jeon, Je-Eon; Seo, Ki-Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The concrete overpack of the cask provides radiation shielding as well as physical protection for inner canister against external mechanical shock. When the overpack undergoes a severe missile impact which might be caused by tornado or aircraft crash, it should sustain minimal level of structural integrity so that the radiation shielding and the retrievability of canister are maintained. Empirical formulas have been developed for the evaluation of concrete damage but those formulas can be used only for local damage evaluation and not for global damage evaluation. In this research, a series of numerical simulations and tests have been performed to evaluate the damage of two types of concrete overpack segment models under high speed missile impact. It is shown that appropriate modeling of material failure is crucial in this kind of analyses and finding the correct failure parameters may not be straightforward. When comparing the simulation results with the test results, it is shown that neither setting, case 1 and 2 provides results with consistent agreement with test results. That is, case 1 setting is more close to reality in Type 1 model analysis, but for Type 2, case 2 setting provides more close results to the reality. In both the case, not enough deformation is predicted by simulation compared to the tests. Weak failure and eroding criteria give larger penetration depth with insufficient overall damage due to energy loss with element erosion.

  15. Creep life simulations of EB welded copper overpack

    Energy Technology Data Exchange (ETDEWEB)

    Holmstroem, S.; Laukkanen, A.; Andersson, T. [VTT Technical Research Centre of Finland, Espoo (Finland)

    2013-12-15

    The long term life predictions of copper overpack (sealed by EB welding in Finland) have previously been based on stress estimations that vary over a wide range, typically between 40-100 MPa. These values are usually not based on structural calculation including the EB-weld that increases the complexity of the stress state in the copper overpack. This report will attempt to pinpoint and simulate the stresses and strains developing in the copper overpack during its first decennia of repository service by advanced FEA simulations including the impact of the EB-weld. The main challenge of this work is the extrapolation of the creep strain response of OFP copper to the service relevant loads and temperatures. The uniaxial creep model is translated to a multiaxial constitutive equation form with adequate computational efficiency. The copper overpack strain and stress evolution has been simulated at up to 100 000 years at a conservative constant temperature of 80 deg C with 14 MPa of external pressure. The results indicate rapid creep relaxation in the initial stages after the load has been applied followed by limited creep strain accumulation thereafter. Local elastic-plastic and creep deformation is predicted at the EB weld root with a total strain of below 12 %. The predicted stresses after external loading and short term relaxation are moderate and the impact of weld residual stresses and the lower creep rupture properties of the EB seem not to be detrimental to the predicted long term creep response. The simulation results imply that the most crucial impact on the creep strain accumulation of the copper overpack is related to the OFP copper primary creep properties. The present study predicts sufficiently low creep strains for a 100 000 years canister life with the conservative assumption at a constant temperature of 80 deg C. However a sensitivity study on the impact of primary creep is strongly recommended due to contradicting analysis results from earlier FEA

  16. IMPACT ANALYSES AND TESTS OF CONCRETE OVERPACKS OF SPENT NUCLEAR FUEL STORAGE CASKS

    Directory of Open Access Journals (Sweden)

    SANGHOON LEE

    2014-02-01

    Full Text Available A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

  17. Assessment of a spent fuel disposal canister. Assessment studies for a copper canister with cast steel inner component

    Energy Technology Data Exchange (ETDEWEB)

    Bond, A.E.; Hoch, A.R.; Jones, G.D.; Tomczyk, A.J.; Wiggin, R.M.; Worraker, W.J. [AEA Technology, Harwell (United Kingdom)

    1997-05-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden, is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in vertical storage holes drilled in a series of caverns excavated from the granite bedrock at a depth of about 500 m. Each canister will be surrounded by compacted bentonite clay. In this report, a simple model of the behaviour of the canister subsequent to a first breach in its copper overpack is developed. This model is used to predict: -the ingress of water to the canister (as a function of the size and the shape of the initial defect, the buffer conductivity, the corrosion rate and the pressure inside the canister); -the build-up of corrosion products in the canister (as a function of the available water in the canister, the corrosion rate and the properties of the corrosion products); -the effect of corrosion on the structural integrity of the canister. A number of different scenarios for the location of the breach in the copper overpack are considered.

  18. Preliminary Transportation, Aging and Disposal Canister System Performance Specification

    Energy Technology Data Exchange (ETDEWEB)

    C.A Kouts

    2006-11-22

    This document provides specifications for selected system components of the Transportation, Aging and Disposal (TAD) canister-based system. A list of system specified components and ancillary components are included in Section 1.2. The TAD canister, in conjunction with specialized overpacks will accomplish a number of functions in the management and disposal of spent nuclear fuel. Some of these functions will be accomplished at purchaser sites where commercial spent nuclear fuel (CSNF) is stored, and some will be performed within the Office of Civilian Radioactive Waste Management (OCRWM) transportation and disposal system. This document contains only those requirements unique to applications within Department of Energy's (DOE's) system. DOE recognizes that TAD canisters may have to perform similar functions at purchaser sites. Requirements to meet reactor functions, such as on-site dry storage, handling, and loading for transportation, are expected to be similar to commercially available canister-based systems. This document is intended to be referenced in the license application for the Monitored Geologic Repository (MGR). As such, the requirements cited herein are needed for TAD system use in OCRWM's disposal system. This document contains specifications for the TAD canister, transportation overpack and aging overpack. The remaining components and equipment that are unique to the OCRWM system or for similar purchaser applications will be supplied by others.

  19. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  20. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  1. Data quality objectives for K West canister sludge sampling

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J., Westinghouse Hanford

    1996-12-11

    Data Quality Objectives have been developed for a limited campaign of sampling K Basin canister sludge. Specifically, samples will be taken from the sealed K West Basin fuel canisters. Characterization of the sludge in these canisters will address the needs of fuel retrieval which are to collect and transport sludge which is currently in the canisters. Data will be gathered on physical properties (such as viscosity, particle size, density, etc.) as well as on chemical and radionuclide constituents and radiation levels of sludge. The primary emphasis will be on determining radionuclide concentrations to be deposited on Ion Exchange Modules (IXMS) during canister opening and fuel retrieval. The data will also be useful in determining whether K West Basin sludge meets the waste acceptance criteria for Hanford waste tanks as a backup disposal concept and these data will also supply information on the properties of sludge material which will1403 accompany fuel elements in the Multi-Canister Overpacks (MCOS) as envisioned in the Integrated Process Strategy (IPS).

  2. Methodology to make a robust estimation of the carbon steel overpack lifetime with respect to the Belgian Supercontainer design

    Science.gov (United States)

    Kursten, B.; Druyts, F.

    2008-09-01

    The Supercontainer (SC) design is the preferred Belgian option for the final disposal of vitrified high-level waste (VHLW) and spent fuel (SF) in deep underground clay layers. The SC consists of a carbon steel overpack, containing VHLW canisters or SF assemblies, surrounded by a thick concrete buffer, which in turn, is entirely encased in a stainless steel envelope. An integrated R&D strategy is developed to demonstrate and defend that the integrity of the carbon steel overpack can be ensured at least during the thermal phase. This integrated approach, proposed to estimate the lifetime of the carbon steel overpack, consists of three steps: lifetime prediction, validation, and confidence building. Under the predicted conditions within the SC (highly alkaline concrete buffer), the carbon steel overpack is expected to undergo uniform corrosion (passive dissolution). The methodology exists in demonstrating that corrosion forms other than uniform corrosion (e.g. localised corrosion such as pitting corrosion, crevice corrosion and stress corrosion cracking) cannot occur ('exclusion principle'). This paper elaborates on how this methodology is implemented.

  3. Materials for Consideration in Standardized Canister Design Activities.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R.; Ilgen, Anastasia Gennadyevna; Enos, David George; Teich-McGoldrick, Stephanie; Hardin, Ernest

    2014-10-01

    This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to

  4. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    B. Gorpani

    2000-06-23

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane hoist, DC loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the DC is fully loaded, the Disposal Container Transport System moves the DC to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister handling

  5. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    CERN Document Server

    Lewis, M E

    2000-01-01

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

  6. Status report, canister fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Emilsson, Goeran [CSM Materialteknik AB, Linkoeping (Sweden)

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  7. HLW Canister and Can-In-Canister Drop Calculation

    Energy Technology Data Exchange (ETDEWEB)

    H. Marr

    1999-09-15

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver.

  8. SLUDGE TREATMENT PROJECT COST COMPARISON BETWEEN HYDRAULIC LOADING AND SMALL CANISTER LOADING CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    GEUTHER J; CONRAD EA; RHOADARMER D

    2009-08-24

    The Sludge Treatment Project (STP) is considering two different concepts for the retrieval, loading, transport and interim storage of the K Basin sludge. The two design concepts under consideration are: (1) Hydraulic Loading Concept - In the hydraulic loading concept, the sludge is retrieved from the Engineered Containers directly into the Sludge Transport and Storage Container (STSC) while located in the STS cask in the modified KW Basin Annex. The sludge is loaded via a series of transfer, settle, decant, and filtration return steps until the STSC sludge transportation limits are met. The STSC is then transported to T Plant and placed in storage arrays in the T Plant canyon cells for interim storage. (2) Small Canister Concept - In the small canister concept, the sludge is transferred from the Engineered Containers (ECs) into a settling vessel. After settling and decanting, the sludge is loaded underwater into small canisters. The small canisters are then transferred to the existing Fuel Transport System (FTS) where they are loaded underwater into the FTS Shielded Transfer Cask (STC). The STC is raised from the basin and placed into the Cask Transfer Overpack (CTO), loaded onto the trailer in the KW Basin Annex for transport to T Plant. At T Plant, the CTO is removed from the transport trailer and placed on the canyon deck. The CTO and STC are opened and the small canisters are removed using the canyon crane and placed into an STSC. The STSC is closed, and placed in storage arrays in the T Plant canyon cells for interim storage. The purpose of the cost estimate is to provide a comparison of the two concepts described.

  9. 42 CFR 438.708 - Termination of an MCO or PCCM contract.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Termination of an MCO or PCCM contract. 438.708... SERVICES (CONTINUED) MEDICAL ASSISTANCE PROGRAMS MANAGED CARE Sanctions § 438.708 Termination of an MCO or PCCM contract. A State has the authority to terminate an MCO or PCCM contract and enroll that...

  10. Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

    Energy Technology Data Exchange (ETDEWEB)

    Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

    2001-04-01

    This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

  11. MCO Membranes: Enhanced Selectivity in High-Flux Class

    Science.gov (United States)

    Boschetti-de-Fierro, Adriana; Voigt, Manuel; Storr, Markus; Krause, Bernd

    2015-12-01

    Novel MCO high-flux membranes for hemodialysis have been developed with optimized permeability, allowing for filtration close to that of the natural kidney. A comprehensive in vitro characterization of the membrane properties by dextran filtration is presented. The sieving profile of pristine membranes, as well as that of membranes exposed to blood for 40 minutes, are described. The effective pore size (Stokes-Einstein radius) was estimated from filtration experiments before and after blood exposure, and results were compared to hydrodynamic radii of middle and large uremic toxins and essential proteins. The results indicate that the tailored pore sizes of the MCO membranes promote removal of large toxins while ensuring the retention of albumin.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-05-27

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

  13. Analysis of K west basin canister gas

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

    Gas and Liquid samples have been collected from a selection of the approximately 3,820 spent fuel storage canisters in the K West Basin. The samples were taken to characterize the contents of the gas and water in the canisters providing source term information for two subprojects of the Spent Nuclear Fuel Project (SNFP) (Fulton 1994): the K Basins Integrated Water Treatment System Subproject (Ball 1996) and the K Basins Fuel Retrieval System Subproject (Waymire 1996). The barrels of ten canisters were sampled for gas and liquid in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results from the first campaign have been reported (Trimble 1995a, 1995b, 1996a, 1996b). The analysis results from the second campaign liquid samples have been documented (Trimble and Welsh 1997; Trimble 1997). This report documents the results for the gas samples from the second campaign and evaluates all gas data in terms of expected releases when opening the canisters for SNFP activities. The fuel storage canisters consist of two closed and sealed barrels, each with a gas trap. The barrels are attached at a trunion to make a canister, but are otherwise independent (Figure 1). Each barrel contains up to seven N Reactor fuel element assemblies. A gas space of nitrogen was established in the top 2.2 to 2.5 inches (5.6 to 6.4 cm) of each barrel. Many of the fuel elements were damaged allowing the metallic uranium fuel to be corroded by the canister water. The corrosion releases fission products and generates hydrogen gas. The released gas mixes with the gas-space gas and excess gas passes through the gas trap into the basin water. The canister design does not allow canister water to be exchanged with basin water.

  14. 42 CFR 435.212 - Individuals who would be ineligible if they were not enrolled in an MCO or PCCM.

    Science.gov (United States)

    2010-10-01

    ... not enrolled in an MCO or PCCM. 435.212 Section 435.212 Public Health CENTERS FOR MEDICARE & MEDICAID... Disabled § 435.212 Individuals who would be ineligible if they were not enrolled in an MCO or PCCM. The State agency may provide that a recipient who is enrolled in an MCO or PCCM and who becomes...

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

    Energy Technology Data Exchange (ETDEWEB)

    McCoy, J.C.

    1997-09-15

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

  16. Remote controlled mover for disposal canister transfer

    Energy Technology Data Exchange (ETDEWEB)

    Suikki, M. [Optimik Oy, Turku (Finland)

    2013-10-15

    This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

  17. Grain boundary corrosion of copper canister material

    Energy Technology Data Exchange (ETDEWEB)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J. [AEA Technology plc, Harwell (United Kingdom)

    2001-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister. The potential for grain boundary corrosion was investigated by exposing copper specimens, which had undergone different heat treatments and hence had different grain sizes, to aerated artificial bentonite-equilibrated groundwater with two concentrations of chloride, for increasing periods of time. The degree of grain boundary corrosion was determined by atomic force microscopy (AFM) and optical microscopy. AFM showed no increase in grain boundary 'ditching' for low chloride groundwater. In high chloride groundwater the surface was covered uniformly with a fine-grained oxide. No increases in oxide thickness were observed. No significant grain boundary attack was observed using optical microscopy either. The work suggests that in aerated artificial groundwaters containing chloride ions, grain boundary corrosion of copper is unlikely to adversely affect SKB's copper canisters.

  18. Drop Calculations of HLW Canister and Pu Can-in-Canister

    Energy Technology Data Exchange (ETDEWEB)

    Sreten Mastilovic

    2001-07-31

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C ), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document.

  19. Evaluation and improvement of the properties of the novel cystine-knot microprotein McoEeTI for oral administration.

    Science.gov (United States)

    Werle, M; Kafedjiiski, K; Kolmar, H; Bernkop-Schnürch, A

    2007-03-06

    Cystine-knot microproteins exhibit several properties that make them highly interesting as scaffolds for oral peptide drug delivery. It was therefore the aim of the study to evaluate the novel clinically relevant cystine-knot microprotein McoEeTI regarding its potential for oral delivery. Additionally, based on the gained results, important features of McoEeTI were improved. Enzymatic degradation was caused by chymotrypsin, trypsin and porcine small intestinal juice whereas McoEeTI was stable towards elastase, membrane bound proteases, pepsin and porcine gastric juice. Only minor McoEeTI degradation was observed during a 24h incubation period in rat plasma. In the presence of various physiological ions about 50% of McoEeTI formed di- and/or trimers. P(app) value of McoEeTI was determined to be (7.4+/-0.4)x10(-6)cm/s. Sodium caprate and polycarbophil-cysteine (PCP-Cys) had no beneficial effect on McoEeTI permeation, whereas the utilization of a chitosan-thiobutylamidine (Chito-TBA) system improved McoEeTI permeation 3-fold. Enzymatic stability could be strongly improved by the utilization of Bowman-Birk-Inhibitor (BBI) as well as PCP-Cys. In conclusion, this study indicates that McoEeTI represents a promising candidate as a novel scaffold for oral peptide drug delivery.

  20. Groundwork for Universal Canister System Development

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Mike [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Prouty, Jeralyn L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Craig, Brian [Argonne National Lab. (ANL), Argonne, IL (United States); Han, Zenghu [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, John Hok [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yung [Argonne National Lab. (ANL), Argonne, IL (United States); Pope, Ron [Argonne National Lab. (ANL), Argonne, IL (United States); Connolly, Kevin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feldman, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jarrell, Josh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Radulescu, Georgeta [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Alan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The mission of the United States Department of Energy's Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and go vernment - sponsored nuclear energy re search. S ome of the waste s that that must be managed have be en identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014 a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister - based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister - based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE's Office of Nuclear Energy Used Fuel Dis position Camp aign's Deep Borehole Field Test . Groundwork for Universal Ca nister System Development September 2015 ii W astes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system

  1. Spent nuclear fuel project product specification

    Energy Technology Data Exchange (ETDEWEB)

    Pajunen, A.L.

    1998-01-30

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification.

  2. Canister storage building hazard analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Krahn, D.E.; Garvin, L.J.

    1997-07-01

    This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the final CSB safety analysis report (SAR) and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Report, and implements the requirements of DOE Order 5480.23, Nuclear Safety Analysis Report.

  3. Telescoping Sample Canister Capture Mechanism (TSCCM)

    Science.gov (United States)

    Kong, Kin Yuen; Gorevan, Stephen; Mukherjee, Suparna; Wilson, Jack

    2003-11-01

    Sample return from solar system bodies including planets, moons, comets and asteroids is of high importance within the space science community. A returned sample will allow much more elaborate and detailed analysis not feasible through remote robotic analysis. For this reason, Honeybee Robotics has developed a low-cost reusable, automated on-orbit sample canister capture mechanism. The purpose of the mechanism is to capture a full sample canister and transfer it to a storage cache, sample return spacecraft, or on-orbit laboratory for further scientific study. The current design allows for reliable misalignment-compensated capture for various sample container geometries in any initial orientation. After capture, the sample canister is aligned and presented for transfer. Honeybee has demonstrated the concept through tests of two- and three-dimensional telescopic capture mechanism breadboards. The telescopic capture mechanism design is scalable, minimizes volume and can be made of lightweight material to minmize mass, all of which are critical aspects of spacecraft design.

  4. Stress corrosion cracking of copper canisters

    Energy Technology Data Exchange (ETDEWEB)

    King, Fraser (Integrity Corrosion Consulting Limited (Canada)); Newman, Roger (Univ. of Toronto (Canada))

    2010-12-15

    A critical review is presented of the possibility of stress corrosion cracking (SCC) of copper canisters in a deep geological repository in the Fennoscandian Shield. Each of the four main mechanisms proposed for the SCC of pure copper are reviewed and the required conditions for cracking compared with the expected environmental and mechanical loading conditions within the repository. Other possible mechanisms are also considered, as are recent studies specifically directed towards the SCC of copper canisters. The aim of the review is to determine if and when during the evolution of the repository environment copper canisters might be susceptible to SCC. Mechanisms that require a degree of oxidation or dissolution are only possible whilst oxidant is present in the repository and then only if other environmental and mechanical loading conditions are satisfied. These constraints are found to limit the period during which the canisters could be susceptible to cracking via film rupture (slip dissolution) or tarnish rupture mechanisms to the first few years after deposition of the canisters, at which time there will be insufficient SCC agent (ammonia, acetate, or nitrite) to support cracking. During the anaerobic phase, the supply of sulphide ions to the free surface will be transport limited by diffusion through the highly compacted bentonite. Therefore, no HS. will enter the crack and cracking by either of these mechanisms during the long term anaerobic phase is not feasible. Cracking via the film-induced cleavage mechanism requires a surface film of specific properties, most often associated with a nano porous structure. Slow rates of dissolution characteristic of processes in the repository will tend to coarsen any nano porous layer. Under some circumstances, a cuprous oxide film could support film-induced cleavage, but there is no evidence that this mechanism would operate in the presence of sulphide during the long-term anaerobic period because copper sulphide

  5. 42 CFR 435.326 - Individuals who would be ineligible if they were not enrolled in an MCO or PCCM.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Individuals who would be ineligible if they were not enrolled in an MCO or PCCM. 435.326 Section 435.326 Public Health CENTERS FOR MEDICARE & MEDICAID... MCO or PCCM. If the agency provides Medicaid to the categorically needy under § 435.212, it...

  6. Computational analysis of the MCoTI-II plant defence knottin reveals a novel intermediate conformation that facilitates trypsin binding

    Science.gov (United States)

    Jones, Peter M.; George, Anthony M.

    2016-03-01

    MCoTI-I and II are plant defence proteins, potent trypsin inhibitors from the bitter gourd Momordica cochinchinensis. They are members of the Knottin Family, which display exceptional stability due to unique topology comprising three interlocked disulfide bridges. Knottins show promise as scaffolds for new drug development. A crystal structure of trypsin-bound MCoTI-II suggested that loop 1, which engages the trypsin active site, would show decreased dynamics in the bound state, an inference at odds with an NMR analysis of MCoTI-I, which revealed increased dynamics of loop 1 in the presence of trypsin. To investigate this question, we performed unrestrained MD simulations of trypsin-bound and free MCoTI-II. This analysis found that loop 1 of MCoTI-II is not more dynamic in the trypsin-bound state than in the free state. However, it revealed an intermediate conformation, transitional between the free and bound MCoTI-II states. The data suggest that MCoTI-II binding involves a process in which initial interaction with trypsin induces transitions between the free and intermediate conformations, and fluctuations between these states account for the increase in dynamics of loop 1 observed for trypsin-bound MCoTI-I. The MD analysis thus revealed new aspects of the inhibitors’ dynamics that may be of utility in drug design.

  7. 42 CFR 438.420 - Continuation of benefits while the MCO or PIHP appeal and the State fair hearing are pending.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Continuation of benefits while the MCO or PIHP... CARE Grievance System § 438.420 Continuation of benefits while the MCO or PIHP appeal and the State... before the later of the following: (1) Within ten days of the MCO or PIHP mailing the notice of...

  8. Medium Cut-Off (MCO) Membranes Reduce Inflammation in Chronic Dialysis Patients—A Randomized Controlled Clinical Trial

    Science.gov (United States)

    Zickler, Daniel; Schindler, Ralf; Willy, Kevin; Martus, Peter; Pawlak, Michael; Storr, Markus; Hulko, Michael; Boehler, Torsten; Glomb, Marcus A.; Liehr, Kristin; Henning, Christian; Templin, Markus; Trojanowicz, Bogusz; Ulrich, Christof; Werner, Kristin; Fiedler, Roman; Girndt, Matthias

    2017-01-01

    Background To increase the removal of middle-sized uremic toxins a new membrane with enhanced permeability and selectivity, called Medium Cut-Off membrane (MCO-Ci) has been developed that at the same time ensures the retention of albumin. Because many middle-sized substances may contribute to micro-inflammation we hypothesized that the use of MCO-Ci influences the inflammatory state in hemodialysis patients. Methods The randomized crossover trial in 48 patients compared MCO-Ci dialysis to High-flux dialysis of 4 weeks duration each plus 8 weeks extension phase. Primary endpoint was the gene expression of TNF-α and IL-6 in peripheral blood mononuclear cells (PBMCs), secondary endpoints were plasma levels of specified inflammatory mediators and cytokines. Results After four weeks of MCO-Ci the expression of TNF-α mRNA (Relative quantification (RQ) from 0.92 ± 0.34 to 0.75 ± 0.31, -18.5%, pkappa and lambda free light chains, urea and an increase for Lp-PLA2 (PLA2G7) compared to High-flux. Albumin levels dropped significantly after 4 weeks of MCO dialysis but increased after additional 8 weeks of MCO dialysis. Twelve weeks treatment with MCO-Ci was well tolerated regarding the number of (S)AEs. In the extension period levels of CRP, TNF-α-mRNA and IL-6 mRNA remained stable in High-flux as well as in MCO-Ci. Conclusions MCO-Ci dialyzers modulate inflammation in chronic HD patients to a greater extent compared to High-flux dialyzers. Transcription of pro-inflammatory cytokines in peripheral leukocytes is markedly reduced and removal of soluble mediators is enhanced with MCO dialysis. Serum albumin concentrations stabilize after an initial drop. These results encourage further trials with longer treatment periods and clinical endpoints. PMID:28085888

  9. Thermal Predictions of the Cooling of Waste Glass Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-11-01

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

  10. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    D.T. Dexheimer

    2004-02-27

    The purpose of this calculation is to estimate radiation doses received by personnel working in the Canister Handling Facility (CHF) performing operations to receive transportation casks, transfer wastes, prepare waste packages, perform associated equipment maintenance. The specific scope of work contained in this calculation covers individual worker group doses on an annual basis, and includes the contributions due to external and internal radiation. The results of this calculation will be used to support the design of the CHF and provide occupational dose estimates for the License Application.

  11. Canister storage building hazard analysis report

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    1999-05-11

    This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the CSB final safety analysis report (FSAR) and documents the results. The hazard analysis was performed in accordance with the DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', and meets the intent of HNF-PRO-704, ''Hazard and Accident Analysis Process''. This hazard analysis implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports''.

  12. EVALUATION OF REQUIREMENTS FOR THE DWPF HIGHER CAPACITY CANISTER

    Energy Technology Data Exchange (ETDEWEB)

    Miller, D.; Estochen, E.; Jordan, J.; Kesterson, M.; Mckeel, C.

    2014-08-05

    The Defense Waste Processing Facility (DWPF) is considering the option to increase canister glass capacity by reducing the wall thickness of the current production canister. This design has been designated as the DWPF Higher Capacity Canister (HCC). A significant decrease in the number of canisters processed during the life of the facility would be achieved if the HCC were implemented leading to a reduced overall reduction in life cycle costs. Prior to implementation of the change, Savannah River National Laboratory (SRNL) was requested to conduct an evaluation of the potential impacts. The specific areas of interest included loading and deformation of the canister during the filling process. Additionally, the effect of the reduced wall thickness on corrosion and material compatibility needed to be addressed. Finally the integrity of the canister during decontamination and other handling steps needed to be determined. The initial request regarding canister fabrication was later addressed in an alternate study. A preliminary review of canister requirements and previous testing was conducted prior to determining the testing approach. Thermal and stress models were developed to predict the forces on the canister during the pouring and cooling process. The thermal model shows the HCC increasing and decreasing in temperature at a slightly faster rate than the original. The HCC is shown to have a 3°F ΔT between the internal and outer surfaces versus a 5°F ΔT for the original design. The stress model indicates strain values ranging from 1.9% to 2.9% for the standard canister and 2.5% to 3.1% for the HCC. These values are dependent on the glass level relative to the thickness transition between the top head and the canister wall. This information, along with field readings, was used to set up environmental test conditions for corrosion studies. Small 304-L canisters were filled with glass and subjected to accelerated environmental testing for 3 months. No evidence of

  13. Cold vacuum drying proof of performance (first article testing) test results

    Energy Technology Data Exchange (ETDEWEB)

    MCCRACKEN, K.J.

    1999-06-23

    This report presents and details the test results of the first of a kind process referred to as Cold Vacuum Drying (CVD). The test results are compiled from several months of testing of the first process equipment skid and ancillary components to de-water and dry Multi-Canister Overpacks (MCO) filled with Spent Nuclear Fuel (SNF). The tests results provide design verifications, equipment validations, model validation data, and establish process parameters.

  14. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  15. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.; PIEPHO, M.G.

    2000-03-23

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  16. Description of DWPF reference waste form and canister

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

    This document describes the reference waste form and canister for the Defense Waste Processing Facility (DWPF). The facility is planned for location at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1983. The reference canister is fabricated of 24-in.-OD 304L stainless steel pipe with a dished bottom, domed head, and lifting and welding flanges on the head neck. The overall canister length is 9 ft 10 in., with a wall thickness of 3/8-in. (schedule 20 pipe). The canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected to ensure that a filled canister with its shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be generally compatible with preliminary assessments of repository requirements. The reference waste form is borosilicate glass containing approximately 28 wt % sludge oxides with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains approximately 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. This composition results in a low average leachability in the waste form of approximately 5 x 10/sup -9/ g/cm/sup 2/-day based on /sup 137/Cs over 365 days in 25/sup 0/C water. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approximately 425 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the Stage 1 and Stage 2 processes. The radionuclide content of the canister is about 150,000 curies, with a radiation level of 2 x 10/sup 4/ rem/hour at 1 cm.

  17. COMSOL Multiphysics Model for HLW Canister Filling

    Energy Technology Data Exchange (ETDEWEB)

    Kesterson, M. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-11

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al2O3 and Na2O can contribute to nepheline (generally NaAlSiO4) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered

  18. Structural Sensitivity of Dry Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Karri, Naveen K.; Adkins, Harold E.; Hanson, Brady D.

    2013-09-27

    This LS-DYNA modeling study evaluated a generic used nuclear fuel vertical dry storage cask system under tip-over, handling drop, and seismic load cases to determine the sensitivity of the canister containment boundary to these loads. The goal was to quantify the expected failure margins to gain insight into what material changes over the extended long-term storage lifetime could have the most influence on the security of the containment boundary. It was determined that the tip-over case offers a strong challenge to the containment boundary, and identifies one significant material knowledge gap, the behavior of welded stainless steel joints under high-strain-rate conditions. High strain rates are expected to increase the material’s effective yield strength and ultimate strength, and may decrease its ductility. Determining and accounting for this behavior could potentially reverse the model prediction of a containment boundary failure at the canister lid weld. It must be emphasized that this predicted containment failure is an artifact of the generic system modeled. Vendor specific designs analyze for cask tip-over and these analyses are reviewed and approved by the Nuclear Regulatory Commission. Another location of sensitivity of the containment boundary is the weld between the base plate and the canister shell. Peak stresses at this location predict plastic strains through the whole thickness of the welded material. This makes the base plate weld an important location for material study. This location is also susceptible to high strain rates, and accurately accounting for the material behavior under these conditions could have a significant effect on the predicted performance of the containment boundary. The handling drop case was largely benign to the containment boundary, with just localized plastic strains predicted on the outer surfaces of wall sections. It would take unusual changes in the handling drop scenario to harm the containment boundary, such as

  19. Radon measurements with charcoal canisters temperature and humidity considerations

    Directory of Open Access Journals (Sweden)

    Živanović Miloš Z.

    2016-01-01

    Full Text Available Radon testing by using open-faced charcoal canisters is a cheap and fast screening method. Many laboratories perform the sampling and measurements according to the United States Environmental Protection Agency method - EPA 520. According to this method, no corrections for temperature are applied and corrections for humidity are based on canister mass gain. The EPA method is practiced in the Vinča Institute of Nuclear Sciences with recycled canisters. In the course of measurements, it was established that the mass gain of the recycled canisters differs from mass gain measured by Environmental Protection Agency in an active atmosphere. In order to quantify and correct these discrepancies, in the laboratory, canisters were exposed for periods of 3 and 4 days between February 2015 and December 2015. Temperature and humidity were monitored continuously and mass gain measured. No significant correlation between mass gain and temperature was found. Based on Environmental Protection Agency calibration data, functional dependence of mass gain on humidity was determined, yielding Environmental Protection Agency mass gain curves. The results of mass gain measurements of recycled canisters were plotted against these curves and a discrepancy confirmed. After correcting the independent variable in the curve equation and calculating the corrected mass gain for recycled canisters, the agreement between measured mass gain and Environmental Protection Agency mass gain curves was attained. [Projekat Ministarstva nauke Republike Srbije, br. III43009: New Technologies for Monitoring and Protection of Environment from Harmful Chemical Substances and Radiation Impact

  20. Radiolysis Model Sensitivity Analysis for a Used Fuel Storage Canister

    Energy Technology Data Exchange (ETDEWEB)

    Wittman, Richard S.

    2013-09-20

    This report fulfills the M3 milestone (M3FT-13PN0810027) to report on a radiolysis computer model analysis that estimates the generation of radiolytic products for a storage canister. The analysis considers radiolysis outside storage canister walls and within the canister fill gas over a possible 300-year lifetime. Previous work relied on estimates based directly on a water radiolysis G-value. This work also includes that effect with the addition of coupled kinetics for 111 reactions for 40 gas species to account for radiolytic-induced chemistry, which includes water recombination and reactions with air.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Simulations of the pipe overpack to compute constitutive model parameters for use in WIPP room closure calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Park, Byoung Yoon; Hansen, Francis D.

    2004-07-01

    The regulatory compliance determination for the Waste Isolation Pilot Plant includes the consideration of room closure. Elements of the geomechanical processes include salt creep, gas generation and mechanical deformation of the waste residing in the rooms. The WIPP was certified as complying with regulatory requirements based in part on the implementation of room closure and material models for the waste. Since the WIPP began receiving waste in 1999, waste packages have been identified that are appreciably more robust than the 55-gallon drums characterized for the initial calculations. The pipe overpack comprises one such waste package. This report develops material model parameters for the pipe overpack containers by using axisymmetrical finite element models. Known material properties and structural dimensions allow well constrained models to be completed for uniaxial, triaxial, and hydrostatic compression of the pipe overpack waste package. These analyses show that the pipe overpack waste package is far more rigid than the originally certified drum. The model parameters developed in this report are used subsequently to evaluate the implications to performance assessment calculations.

  3. Design analysis report for the canister

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, Heikki (VTT (Finland)); Sandstroem, Rolf (Materials Science and Engineering, Royal Inst. of Technology, Stockholm (Sweden)); Ryden, Haakan; Johansson, Magnus (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden))

    2010-04-15

    The mechanical strength of the canister (BWR and PWR types) has been studied. The loading processes are taken from the design premises report and some of them, especially the uneven bentonite swelling cases, are further developed in this study and in its references. The canister geometry is described in detail including the manufacturing tolerances of the dimensions. The canister material properties are summarised and the wide material testing programmes and model developments are referenced. The combination of various load cases are rationalised and the conservative combinations are defined. Also the probabilities of various load cases and combinations are assessed for setting reasonable safety margins. The safety margins are used according to ASME Code principles for safety class 1 components. The governing load cases are analysed with 2D- or global 3D-finite-element models including large deformation and non-linear material modelling and, in some cases, also creep. The integrity assessments are partly made from the stress and strain results using global models and partly from fracture resistance analyses using the sub-modelling technique. The sub-model analyses utilize the deformations from the global analyses as constraints on the sub-model boundaries and more detailed finite-element meshes are defined with defects included in the models together with elastic-plastic material models. The J-integral is used as the fracture parameter for the postulated defects. The allowable defect sizes are determined using the measured fracture resistance curves of the insert iron as a reference with respective safety factors according to the ASME Pressure Vessel Code requirements. Based on the BWR canister analyses, the following conclusions can be drawn. The 45 MPa isostatic pressure load case shows very robust and distinct results in that the risk for local collapse is vanishingly small. The probabilistic analysis of plastic collapse only considers the initial local collapse

  4. Deflection measurements of LABAN canister sections in horizontal attitude

    Energy Technology Data Exchange (ETDEWEB)

    Wakeman, W.

    1985-01-08

    Deflection measurements made on the LABAN canister sections indicate that the apparent stiffness of its frames, with all the diagnostics experiments installed, is not significantly different from the stiffness of the bare frames.

  5. Spent nuclear fuel canister storage building conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  6. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    1999-09-09

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  7. Criticality safety evaluation report for the Cold Vacuum Drying Facility`s process water handling system

    Energy Technology Data Exchange (ETDEWEB)

    Roblyer, S.D.

    1998-02-12

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility (CVDF). The controls and limitations on equipment design and operations to control potential criticality occurrences are identified. The effectiveness of equipment design and operation controls in preventing criticality occurrences during normal and abnormal conditions is evaluated and documented in this report. Spent nuclear fuel (SNF) is removed from existing canisters in both the K East and K West Basins and loaded into a multicanister overpack (MCO) in the K Basin pool. The MCO is housed in a shipping cask surrounded by clean water in the annulus between the exterior of the MCO and the interior of the shipping cask. The fuel consists of spent N Reactor and some single pass reactor fuel. The MCO is transported to the CVDF near the K Basins to remove process water from the MCO interior and from the shipping cask annulus. After the bulk water is removed from the MCO, any remaining free liquid is removed by drawing a vacuum on the MCO`s interior. After cold vacuum drying is completed, the MCO is filled with an inert cover gas, the lid is replaced on the shipping cask, and the MCO is transported to the Canister Storage Building. The process water removed from the MCO contains fissionable materials from metallic uranium corrosion. The process water from the MCO is first collected in a geometrically safe process water conditioning receiver tank. The process water in the process water conditioning receiver tank is tested, then filtered, demineralized, and collected in the storage tank. The process water is finally removed from the storage tank and transported from the CVDF by truck.

  8. Remote Welding, NDE and Repair of DOE Standardized Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Eric Larsen; Art Watkins; Timothy R. McJunkin; Dave Pace; Rodney Bitsoi

    2006-05-01

    The U.S. Department of Energy (DOE) created the National Spent Nuclear Fuel Program (NSNFP) to manage DOE’s spent nuclear fuel (SNF). One of the NSNFP’s tasks is to prepare spent nuclear fuel for storage, transportation, and disposal at the national repository. As part of this effort, the NSNFP developed a standardized canister for interim storage and transportation of SNF. These canisters will be built and sealed to American Society of Mechanical Engineers (ASME) Section III, Division 3 requirements. Packaging SNF usually is a three-step process: canister loading, closure welding, and closure weld verification. After loading SNF into the canisters, the canisters must be seal welded and the welds verified using a combination of visual, surface eddy current, and ultrasonic inspection or examination techniques. If unacceptable defects in the weld are detected, the defective sections of weld must be removed, re-welded, and re-inspected. Due to the high contamination and/or radiation fields involved with this process, all of these functions must be performed remotely in a hot cell. The prototype apparatus to perform these functions is a floor-mounted carousel that encircles the loaded canister; three stations perform the functions of welding, inspecting, and repairing the seal welds. A welding operator monitors and controls these functions remotely via a workstation located outside the hot cell. The discussion describes the hardware and software that have been developed and the results of testing that has been done to date.

  9. Impact analysis of stainless steel spent fuel canisters

    Energy Technology Data Exchange (ETDEWEB)

    Aramayo, G.A. [Oak Ridge National Lab., TN (United States); Turner, D.W. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States). Waste Management Organization

    1998-04-01

    This paper presents the results of the numerical analysis performed to asses the structural integrity of spent nuclear fuel (SNF) stainless steel canisters when subjected to impact loads associated with free gravity drops from heights not exceeding 20 ft. The SNF canisters are to be used for the Shipment of radioactive material from the Oak Ridge National Laboratory (ORNL) Site to the Idaho National Engineering and Environmental Laboratory (INEEL) for storage. The Idaho chemical Processing Plant Fuel Receipt Criteria Questionnaire requires that the vertical drop accidents from two heights be analyze. These heights are those that are considered to be critical at the time of unloading the canisters from the shipping cask. The configurations analyzed include a maximum payload of 90 lbs dropping from heights of 20 and 3 ft. The nominal weight of the canister is 23.3 lbs. The analysis has been performed using finite element methods. Innovative analysis techniques are used to capture the effects of failure and separation of canister components. The structural integrity is evaluated in terms of physical deformation and separation of the canister components that may result from failure of components at selected interfaces.

  10. GIBNE canister: a comprehensive analytical and experimental evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard, M.A.

    1983-01-17

    The finite-element computer program GEMINI was used to efficiently and accurately characterize the GIBNE 86 in canister. GEMINI accurately calculated the GIBNE bare frame deflections for all four load cases. The center bulkhead of the 45 ft long cnaister deflected .323 in. when the canister was horizontally supported by its end bulkheads. Several large lead weights were used to simulate (but not accurately characterize) the addition of internal hardware to the canister. The devlection increased to .512 in. with the addition of 8000 lbs to bulkhead 5. With the 8000 lb load moved to bulkhead 4 and 8260 lbs added to bulkhead 6, the deflection increased to .678 in. Deflections calculated by GEMINI were conservative by 3 to 5%. GEMINI correctly predicted the stress distribution in the bare frame cable trays. The GIBNE tests and analyses accurately characterized the GIBNE bare frame. However, the experimental results did not separate individual effects of the lines-of-sight and end fixtures. As a result, the numerical model can not be validated for a canister including lines-of-sight. During calendar year 1983, the LABAN test will characterize the individual effects of the lines-of-sight and the end fixtures. At that time the numerical model will be fine-tuned to match the experimental results. We will then be able to analytically predict canister alignment changes under a wide variety of loading conditions.

  11. Design, production and initial state of the canister

    Energy Technology Data Exchange (ETDEWEB)

    Cederqvist, Lars; Johansson, Magnus; Leskinen, Nina; Ronneteg, Ulf

    2010-12-15

    The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility.The report provides input on the initial state of the canisters to the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides input to the operational safety report, SR-Operation, on how the canisters shall be handled and disposed. The report presents the design premises and reference design of the canister and verifies the conformity of the reference design to the design premises. The production methods and the ability to produce canisters according to the reference design are described. Finally, the initial state of the canisters and their conformity to the reference design and design premises are presented

  12. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Enos, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  13. Debris Removal Project K West Canister Cleaning System Performance Specification

    Energy Technology Data Exchange (ETDEWEB)

    FARWICK, C.C.

    1999-12-09

    Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. Design criteria for a Canister Cleaning System to be installed in the KW Basin. This documents the requirements for design and installation of the system.

  14. Evaluation of the Frequencies for Canister Inspections for SCC

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-02-02

    This report fulfills the M3 milestone M3FT-15SN0802042, “Evaluate the Frequencies for Canister Inspections for SCC” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. It reviews the current state of knowledge on the potential for stress corrosion cracking (SCC) of dry storage canisters and evaluates the implications of this state of knowledge on the establishment of an SCC inspection frequency. Models for the prediction of SCC by the Japanese Central Research Institute of Electric Power Industry (CRIEPI), the United States (U.S.) Electric Power Research Institute (EPRI), and Sandia National Laboratories (SNL) are summarized, and their limitations discussed.

  15. SLUDGE TREATMENT PROJECT KOP CONCEPTUAL DESIGN CONTROL DECISION REPORT

    Energy Technology Data Exchange (ETDEWEB)

    CARRO CA

    2010-03-09

    This control decision addresses the Knock-Out Pot (KOP) Disposition KOP Processing System (KPS) conceptual design. The KPS functions to (1) retrieve KOP material from canisters, (2) remove particles less than 600 {micro}m in size and low density materials from the KOP material, (3) load the KOP material into Multi-Canister Overpack (MCO) baskets, and (4) stage the MCO baskets for subsequent loading into MCOs. Hazard and accident analyses of the KPS conceptual design have been performed to incorporate safety into the design process. The hazard analysis is documented in PRC-STP-00098, Knock-Out Pot Disposition Project Conceptual Design Hazard Analysis. The accident analysis is documented in PRC-STP-CN-N-00167, Knock-Out Pot Disposition Sub-Project Canister Over Lift Accident Analysis. Based on the results of these analyses, and analyses performed in support of MCO transportation and MCO processing and storage activities at the Cold Vacuum Drying Facility (CVDF) and Canister Storage Building (CSB), control decision meetings were held to determine the controls required to protect onsite and offsite receptors and facility workers. At the conceptual design stage, these controls are primarily defined by their safety functions. Safety significant structures, systems, and components (SSCs) that could provide the identified safety functions have been selected for the conceptual design. It is anticipated that some safety SSCs identified herein will be reclassified based on hazard and accident analyses performed in support of preliminary and detailed design.

  16. Estimation of CANDU spent fuel disposal canister lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Lee, Min Soo; Hwang, Yong Soo; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Active nuclear energy utilization causes significant spent fuel accumulation problem. The cumulative amount of spent fuel is about 10,083 ton as of Dec. 2008, and is expected to increase up to 19,000 ton by 2020. Of those, CANDU spent fuels account for more than 60% of the total amounts. CANDU spent fuels had been stored in dry concrete silos since 1991 and during the past 15 years, 300 silos were constructed and {approx}3,200 ton of spent fuels are stored now. Another dry storage facility MACSTOR /KN-400 will store new-coming CANDU spent fuels from 2009. But, after intermediate storage ends, all CANDU spent fuels have to be disposed within multi-layer metallic canister which is composed of cast iron inside and copper outside. Canister lifetime estimation, therefore, is very important for the final disposal safety analysis. The most significant factor of lifetime is copper corrosion, and Y. S. Hwang developed a corrosion model in order to predict the general corrosion effect on copper canister lifetime during the final disposal period. This research applied his model to KURT1 where many disposal researches are being performed actively and the results shows safe margin of the copper canister for the very long-term disposal.

  17. OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains & Science which has articles on the Yucca Mountain project. The OCRWM provides information about OCRWM activities and in this issue has articles on multi-purpose canister design, and transportation cask trailer.

  18. Storage and disposal of radioactive waste as glass in canisters

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E.

    1978-12-01

    A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal.

  19. Canister Cleaning System Final Design Report Project A-2A

    Energy Technology Data Exchange (ETDEWEB)

    FARWICK, C.C.

    2000-06-15

    Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. The Canister Cleaning System (CCS) is part of the Debris Removal Project. The CCS will be installed in the KW Basin and operated during the fuel removal activity. The KW Basin has approximately 3600 canisters that require removal from the basin. The CCS is being designed to ''clean'' empty fuel canisters and lids and package them for disposal to the Environmental Restoration Disposal Facility complex. The system will interface with the KW Basin and be located in the Dummy Elevator Pit.

  20. The development of a Martian atmospheric Sample collection canister

    Science.gov (United States)

    Kulczycki, E.; Galey, C.; Kennedy, B.; Budney, C.; Bame, D.; Van Schilfgaarde, R.; Aisen, N.; Townsend, J.; Younse, P.; Piacentine, J.

    The collection of an atmospheric sample from Mars would provide significant insight to the understanding of the elemental composition and sub-surface out-gassing rates of noble gases. A team of engineers at the Jet Propulsion Laboratory (JPL), California Institute of Technology have developed an atmospheric sample collection canister for Martian application. The engineering strategy has two basic elements: first, to collect two separately sealed 50 cubic centimeter unpressurized atmospheric samples with minimal sensing and actuation in a self contained pressure vessel; and second, to package this atmospheric sample canister in such a way that it can be easily integrated into the orbiting sample capsule for collection and return to Earth. Sample collection and integrity are demonstrated by emulating the atmospheric collection portion of the Mars Sample Return mission on a compressed timeline. The test results achieved by varying the pressure inside of a thermal vacuum chamber while opening and closing the valve on the sample canister at Mars ambient pressure. A commercial off-the-shelf medical grade micro-valve is utilized in the first iteration of this design to enable rapid testing of the system. The valve has been independently leak tested at JPL to quantify and separate the leak rates associated with the canister. The results are factored in to an overall system design that quantifies mass, power, and sensing requirements for a Martian atmospheric Sample Collection (MASC) canister as outlined in the Mars Sample Return mission profile. Qualitative results include the selection of materials to minimize sample contamination, preliminary science requirements, priorities in sample composition, flight valve selection criteria, a storyboard from sample collection to loading in the orbiting sample capsule, and contributions to maintaining “ Earth” clean exterior surfaces on the orbiting sample capsule.

  1. Defects which might occur in the copper-iron canister classified according to their likely effect on canister integrity

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom)

    2000-06-15

    Earlier studies identified the material and manufacturing defects that might occur in serially produced canisters to the SKB reference design. This study has considered the defects, which were identified in the earlier works and classified them in terms of their importance to the durability of the canister in service. It has depended on, observations made by the writer over a seven-year involvement with SKI, literature studies and consultation with experts. For ease of reference each section of the report contains a table which includes information on defects taken from the earlier work plus the classification arising from this work. A study has been conducted to identify the material and manufacturing defects that might occur in serially produced canisters to the SKB reference design. The study has depended on cooperation of contractors engaged by SKB to participate in the development program, SKB staff, observations made by the writer over a five-year involvement with SKI, literature studies and consultation with experts. The candidate manufacturing procedures have been described inasmuch as it has been necessary to do so to make the points related to defects. Where possible, the cause of defects, their likely effects on manufacturing procedures or on durability of the canister and the methods available for their detection are given. For ease of reference each section of the report contains a table which summarises the information in it and, in the final section of the report, all the tables are presented en-bloc.

  2. Water soluble decontamination coating for Defense Waste Processing Facility (DWPF) canisters

    Energy Technology Data Exchange (ETDEWEB)

    Selby, C.L.

    1986-12-17

    Water soluble sodium borate glass coating was successfully codeveloped by Clemson University (Dr. H.G. Lefort) and Du Pont as an alternative decontamination process to frit slurry blasting of Defense Waste Processing Facility (DWPF) canisters. Slurry blasting requires transport of abrasive slurries, might cause galling by entrapped frit particles, and could result in frit slurry freezeup in pumps and retention basins. Contamination can be removed from precoated canisters with a gentle hot water rinse. Glass waste spilled on a coated canister will spall off spontaneously during canister cooling. A glass coating appears to prevent transfer of contamination to the Canister Decontamination Cell (CDC) guides and cradle. 1 ref., 5 tabs.

  3. Design of a MCoTI-Based Cyclotide with Angiotensin (1-7-Like Activity

    Directory of Open Access Journals (Sweden)

    Teshome Aboye

    2016-01-01

    Full Text Available We report for the first time the design and synthesis of a novel cyclotide able to activate the unique receptor of angiotensin (1-7 (AT1-7, the MAS1 receptor. This was accomplished by grafting an AT1-7 peptide analog onto loop 6 of cyclotide MCoTI-I using isopeptide bonds to preserve the α-amino and C-terminal carboxylate groups of AT1-7, which are required for activity. The resulting cyclotide construct was able to adopt a cyclotide-like conformation and showed similar activity to that of AT1-7. This cyclotide also showed high stability in human serum thereby providing a promising lead compound for the design of a novel type of peptide-based in the treatment of cancer and myocardial infarction.

  4. Evaluation of helium impurity impacts on Spent Nuclear Fuel project processes (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    SHERRELL, D.L.

    1999-09-21

    This document identifies the types and quantities of impurities that may be present within helium that is introduced into multi-canister overpacks (MCO)s by various SNF Project facilities, including, but not limited to the Cold Vacuum Drying (CVD) Facility (CVDF). It then evaluates possible impacts of worst case impurity inventories on MCO drying, transportation, and storage processes. Based on the evaluation results, this document: (1) concludes that the SNF Project helium procurement specification can be a factor-of-ten less restrictive than a typical vendor's standard offering (99.96% pure versus the vendor's 99.997% pure standard offering); (2) concludes that the CVDF's current 99.5% purity requirement is adequate to control the quality of the helium that is delivered to the MCO by the plant's helium distribution system; and (3) recommends specific impurity limits for both of the above cases.

  5. Qualification of final closure for disposal container II - applicability of TOFD and phased array technique for overpack welding

    Energy Technology Data Exchange (ETDEWEB)

    Asano, H.; Kawahara, K. [Radioactive Waste Management Funding and Research Center (RWMC) (Japan); Arakawa, T. [Ishikawajima-Harima Heavy Industries Co. Ltd. (Japan); Kurokawa, M. [Mitsubishi Heavy Industries Ltd. (Japan)

    2002-07-01

    With a focus on carbon steel, which is one of the candidate materials for the disposal container used in the geological disposal of high-level radioactive waste in Japan, the defect detection capabilities were examined regarding engineering defects of the TOFD technique, an ultrasonic testing method, and the phased array TOFD technique as non-destructive test techniques for the inspection of the weld of a carbon steel overpack. Regarding the TOFD technique, a measurement was conducted concerning the influence of the crossing angle of the ultrasonic beams on the capability of detect flaws, for examining the detection characteristics of the technique in relation to the lid structure of an overpack, and it was pointed out that it is appropriate to consider the lower tip of slit as the reference flaw. Based on the measurements and calculations regarding sound pressure distribution, projections about the scope covered by one test session were made and the optimum testing conditions were examined. Regarding the phased array TOFP technique, the detectability and quantification characteristics were investigated, and comparisons with those of the TOFD technique and the phased array UT technique were made. From the viewpoint of securing long-term corrosion resistance for an overpack, the ways of thinking for ensuring the quality and long-term integrity of the final sealing area of a disposal container were examined. This study stresses that identifying and defining the defects that are harmful to corrosion allowance is important as well as achieving improvements in the welding and testing techniques, and that the question to solve in particular from now on is how to establish effective means to detect defects on the weld surface and the near surface and how to approach the level of tolerance concerning the defects on and near the surface. (orig.)

  6. Biological Research in Canisters (BRIC) - Light Emitting Diode (LED)

    Science.gov (United States)

    Levine, Howard G.; Caron, Allison

    2016-01-01

    The Biological Research in Canisters - LED (BRIC-LED) is a biological research system that is being designed to complement the capabilities of the existing BRIC-Petri Dish Fixation Unit (PDFU) for the Space Life and Physical Sciences (SLPS) Program. A diverse range of organisms can be supported, including plant seedlings, callus cultures, Caenorhabditis elegans, microbes, and others. In the event of a launch scrub, the entire assembly can be replaced with an identical back-up unit containing freshly loaded specimens.

  7. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  8. PAUT inspection of copper canister: Structural attenuation and POD formulation

    Science.gov (United States)

    Gianneo, A.; Carboni, M.; Mueller, C.; Ronneteg, U.

    2016-02-01

    For inspection of thick-walled (50mm) copper canisters for final disposal of spent nuclear fuel in Sweden, ultrasonic inspection using phased array technique (PAUT) is applied. Because thick-walled copper is not commonly used as a structural material, previous experience on Phased Array Ultrasonic Testing for this type of application is limited. The paper presents the progress in understanding the amplitudes and attenuation changes acting on the Phased Array Ultrasonic Testing inspection of copper canisters. Previous studies showed the existence of a low pass filtering effect and a heterogeneous grain size distribution along the depth, thus affecting both the detectability of defects and their "Probability of Detection" determination. Consequently, the difference between the first and second back wall echoes were not sufficient to determine the local attenuation (within the inspection range), which affects the signal response for each individual defect. Experimental evaluation of structural attenuation was carried out onto step-wedge samples cut from full-size, extruded and pierced & drawn, copper canisters. Effective attenuation values has been implemented in numerical simulations to achieve a Multi Parameter Probability of Detection and to formulate a Model Assisted Probability of Detection through a Monte-Carlo extraction model.

  9. NDE to Manage Atmospheric SCC in Canisters for Dry Storage of Spent Fuel: An Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cuta, Judith M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, Aaron A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-01

    This report documents efforts to assess representative horizontal (Transuclear NUHOMS®) and vertical (Holtec HI-STORM) storage systems for the implementation of non-destructive examination (NDE) methods or techniques to manage atmospheric stress corrosion cracking (SCC) in canisters for dry storage of used nuclear fuel. The assessment is conducted by assessing accessibility and deployment, environmental compatibility, and applicability of NDE methods. A recommendation of this assessment is to focus on bulk ultrasonic and eddy current techniques for direct canister monitoring of atmospheric SCC. This assessment also highlights canister regions that may be most vulnerable to atmospheric SCC to guide the use of bulk ultrasonic and eddy current examinations. An assessment of accessibility also identifies canister regions that are easiest and more difficult to access through the ventilation paths of the concrete shielding modules. A conceivable sampling strategy for canister inspections is to sample only the easiest to access portions of vulnerable regions. There are aspects to performing an NDE inspection of dry canister storage system (DCSS) canisters for atmospheric SCC that have not been addressed in previous performance studies. These aspects provide the basis for recommendations of future efforts to determine the capability and performance of eddy current and bulk ultrasonic examinations for atmospheric SCC in DCSS canisters. Finally, other important areas of investigation are identified including the development of instrumented surveillance specimens to identify when conditions are conducive for atmospheric SCC, characterization of atmospheric SCC morphology, and an assessment of air flow patterns over canister surfaces and their influence on chloride deposition.

  10. LABAN emplacement pipe load-release test and stemming/canister alignment study

    Energy Technology Data Exchange (ETDEWEB)

    Howard, D.L.

    1983-12-02

    An Emplacement Pipe Load-Release Test and a study of downhole alignment during stemming were performed on the LABAN event. The purpose of these experiments was to determine canister and line of sight (LOS) distortion induced by downhole stemming and load-release procedures. The load-release test was aborted at approximately 40% completion due to excessive canister distortions. This report summarizes test results in terms of emplacement pipe loads vs vertical canister motions, canister and LOS lateral displacements, and the changes in LOS alignment that resulted from the downhole stemming and load-release processes.

  11. The Unity connecting module is moved to payload canister

    Science.gov (United States)

    1998-01-01

    In the Space Station Processing Facility, an overhead crane moves the Unity connecting module to the payload canister for transfer to the launch pad. Part of the International Space Station (ISS), Unity is scheduled for launch aboard Space Shuttle Endeavour on Mission STS-88 in December. The Unity is a connecting passageway to the living and working areas of ISS. While on orbit, the flight crew will deploy Unity from the payload bay and attach Unity to the Russian-built Zarya control module which will be in orbit at that time.

  12. The Unity connecting module is moved to payload canister

    Science.gov (United States)

    1998-01-01

    In the Space Station Processing Facility, workers attach the overhead crane that will lift the Unity connecting module from its workstand to move the module to the payload canister. Part of the International Space Station (ISS), Unity is scheduled for launch aboard Space Shuttle Endeavour on Mission STS-88 in December. The Unity is a connecting passageway to the living and working areas of ISS. While on orbit, the flight crew will deploy Unity from the payload bay and attach Unity to the Russian-built Zarya control module which will be in orbit at that time.

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  14. Microwave Temperature Profiler Mounted in a Standard Airborne Research Canister

    Science.gov (United States)

    Mahoney, Michael J.; Denning, Richard F.; Fox, Jack

    2009-01-01

    Many atmospheric research aircraft use a standard canister design to mount instruments, as this significantly facilitates their electrical and mechanical integration and thereby reduces cost. Based on more than 30 years of airborne science experience with the Microwave Temperature Profiler (MTP), the MTP has been repackaged with state-of-the-art electronics and other design improvements to fly in one of these standard canisters. All of the controlling electronics are integrated on a single 4 5-in. (.10 13- cm) multi-layer PCB (printed circuit board) with surface-mount hardware. Improved circuit design, including a self-calibrating RTD (resistive temperature detector) multiplexer, was implemented in order to reduce the size and mass of the electronics while providing increased capability. A new microcontroller-based temperature controller board was designed, providing better control with fewer components. Five such boards are used to provide local control of the temperature in various areas of the instrument, improving radiometric performance. The new stepper motor has an embedded controller eliminating the need for a separate controller board. The reference target is heated to avoid possible emissivity (and hence calibration) changes due to moisture contamination in humid environments, as well as avoiding issues with ambient targets during ascent and descent. The radiometer is a double-sideband heterodyne receiver tuned sequentially to individual oxygen emission lines near 60 GHz, with the line selection and intermediate frequency bandwidths chosen to accommodate the altitude range of the aircraft and mission.

  15. Measurements of Fundamental Fluid Physics of SNF Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Condie, Keith Glenn; Mc Creery, Glenn Ernest; McEligot, Donald Marinus

    2001-09-01

    With the University of Idaho, Ohio State University and Clarksean Associates, this research program has the long-term goal to develop reliable predictive techniques for the energy, mass and momentum transfer plus chemical reactions in drying / passivation (surface oxidation) operations in the transfer and storage of spent nuclear fuel (SNF) from wet to dry storage. Such techniques are needed to assist in design of future transfer and storage systems, prediction of the performance of existing and proposed systems and safety (re)evaluation of systems as necessary at later dates. Many fuel element geometries and configurations are accommodated in the storage of spent nuclear fuel. Consequently, there is no one generic fuel element / assembly, storage basket or canister and, therefore, no single generic fuel storage configuration. One can, however, identify generic flow phenomena or processes which may be present during drying or passivation in SNF canisters. The objective of the INEEL tasks was to obtain fundamental measurements of these flow processes in appropriate parameter ranges.

  16. Thermal-hydraulic assessment of concrete storage cubicle with horizontal 3013 canisters

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-08

    The FIDAP computer code was used to perform a series of analyses to assess the thermal-hydraulic performance characteristics of the concrete plutonium storage cubicles, as modified for the horizontal placement of 3013 canisters. Four separate models were developed ranging from a full height model of the storage cubicle to a very detailed standalone model of a horizontal 3013 canister.

  17. Two-dimensional model of a Space Station Freedom thermal energy storage canister

    Science.gov (United States)

    Kerslake, Thomas W.; Ibrahim, Mounir B.

    1990-01-01

    The Solar Dynamic Power Module being developed for Space Station Freedom uses a eutectic mixture of LiF-CaF2 phase change salt contained in toroidal canisters for thermal energy storage. Results are presented from heat transfer analyses of the phase change salt containment canister. A 2-D, axisymmetric finite difference computer program which models the canister walls, salt, void, and heat engine working fluid coolant was developed. Analyses included effects of conduction in canister walls and solid salt, conduction and free convection in liquid salt, conduction and radiation across salt vapor filled void regions and forced convection in the heat engine working fluid. Void shape, location, growth or shrinkage (due to density difference between the solid and liquid salt phases) were prescribed based on engineering judgement. The salt phase change process was modeled using the enthalpy method. Discussion of results focuses on the role of free-convection in the liquid salt on canister heat transfer performance. This role is shown to be important for interpreting the relationship between ground based canister performance (in l-g) and expected on-orbit performance (in micro-g). Attention is also focused on the influence of void heat transfer on canister wall temperature distributions. The large thermal resistance of void regions is shown to accentuate canister hot spots and temperature gradients.

  18. Commercial radioactive waste management system feasibility with the universal canister concept. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Morissette, R.P.; Schneringer, P.E.; Lane, R.K.; Moore, R.L.; Young, K.A.

    1986-01-01

    A Program Research and Development Announcement (PRDA) was initiated by DOE to solicit from industry new and novel ideas for improvements in the nuclear waste management system. GA Technologies Inc. was contracted to study a system utilizing a universal canister which could be loaded at the reactor and used throughout the waste management system. The proposed canister was developed with the objective of meeting the mission requirements with maximum flexibility and at minimum cost. Canister criteria were selected from a thorough analysis of the spent fuel inventory, and canister concepts were evaluated along with the shipping and storage casks to determine the maximum payload. Engineering analyses were performed on various cask/canister combinations. One important criterion was the interchangeability of the canisters between truck and rail cask systems. A canister was selected which could hold three PWR intact fuel elements or up to eight consolidated PWR fuel elements. One canister could be shipped in an overweight truck cask or six in a rail cask. Economic analysis showed a cost savings of the reference system under consideration at that time.

  19. Aespoe Hard Rock Laboratory Canister Retrieval Test. Microorganisms in buffer from the Canister Retrieval Test - numbers and metabolic diversity

    Energy Technology Data Exchange (ETDEWEB)

    Lydmark, Sara; Pedersen, Karsten (Microbial Analytics Sweden AB (Sweden))

    2011-03-15

    'Canister Retrieval Test' (CRT) is an experiment that started at Aespoe Hard Rock Laboratory (HRL) 2000. CRT is a part of the investigations which evaluate a possible KBS-3 storage of nuclear waste. The primary aim was to see whether it is possible or not to retrieve a copper canister after storage under authentic KBS-3 conditions. However, CRT also provided a unique opportunity to investigate if bacteria survived in the bentonite buffer during storage. Therefore, in connection to the retrieval of the canister microbiological samples were extracted from the bentonite buffer and the bacterial composition was studied. In this report, microbiological analyses of a total of 66 samples at the C2, R10, R9 and R6 levels in the bentonite from CRT are presented and discussed. By culturing bacteria from the bentonite in specific media the following bacterial parameters were investigated: The total amount of culturable heterotrophic aerobic bacteria, sulphate-reducing bacteria, and bacteria that produce the organic compound acetate (acetogens). The biovolume in the bentonite was determined by detection of the ATP content. In addition, bacteria from the bentonite were cultured in different sulphate-reducing media. In these cultures, the presence of the biotic compounds sulphide and acetate was investigated, since these have potentially negative effect on the copper canister in a KBS-3 repository. The results were to some extent compared to density, water content, and temperature data provided by Clay Technology AB. The results showed that 100-102 viable sulphate-reducing and acetogenic bacteria and 102-104 heterotrophic aerobic bacteria g-1 bentonite were present after five years of storage in the rock. Bacteria with several morphologies could be found in the cultures with bentonite. The most bacteria were detected in the bentonite buffer close to the rock but in a few samples also in bentonite close to the copper canister. When the presence of bacteria in the

  20. Reliability in sealing of canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ronneteg, Ulf [Bodycote Materials Testing AB, Nykoeping (Sweden); Cederqvist, Lars; Ryden, Haakan [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Oeberg, Tomas [Tomas Oeberg Konsult AB, Karlskrona (Sweden); Mueller, Christina [Federal Inst. for Materials Research and Testing, Berlin (Germany)

    2006-06-15

    The reliability of the system for sealing the canister and inspecting the weld that has been developed for the Encapsulation plant was investigated. In the investigation the occurrence of discontinuities that can be formed in the welds was determined both qualitatively and quantitatively. The probability that these discontinuities can be detected by nondestructive testing (NDT) was also studied. The friction stir welding (FSW) process was verified in several steps. The variables in the welding process that determine weld quality were identified during the development work. In order to establish the limits within which they can be allowed to vary, a screening experiment was performed where the different process settings were tested according to a given design. In the next step the optimal process setting was determined by means of a response surface experiment, whereby the sensitivity of the process to different variable changes was studied. Based on the optimal process setting, the process window was defined, i.e. the limits within which the welding variables must lie in order for the process to produce the desired result. Finally, the process was evaluated during a demonstration series of 20 sealing welds which were carried out under production-like conditions. Conditions for the formation of discontinuities in welding were investigated. The investigations show that the occurrence of discontinuities is dependent on the welding variables. Discontinuities that can arise were classified and described with respect to characteristics, occurrence, cause and preventive measures. To ensure that testing of the welds has been done with sufficient reliability, the probability of detection (POD) of discontinuities by NDT and the accuracy of size determination by NDT were determined. In the evaluation of the demonstration series, which comprised 20 welds, a statistical method based on the generalized extreme value distribution was fitted to the size estimate of the indications

  1. Study of the consequences of secondary water radiolysis within and surrounding a defective canister

    Energy Technology Data Exchange (ETDEWEB)

    Jinsong Liu; Neretnieks, I. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Chemical Engineering and Technology; Stroemberg, Bo [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2000-11-01

    Consequences of secondary water radiolysis, caused by dispersed radionuclides released from spent nuclear fuel, both inside a defective canister and in the bentonite buffer surrounding the canister have been studied. The dissolution rate of the spent fuel is assumed to be controlled by chemical kinetics. Several cases have been addressed. First a simple mass balance model is presented. Some very conservative assumptions like complete failure of the canister one thousand years after its deposition in the repository and instantaneous oxidation rate of the spent fuel are deliberately made, to explore the upper bound limit of the effect of the secondary water radiolysis on the spent fuel dissolution. The model results show that the spent fuel could possibly be oxidised in an ever-increasing rate with these very simplified assumptions. More realistic and less conservative cases are then considered. In these cases, the canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The spent fuel dissolution is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. The cases are modelled with progressive complication. In the first case the effect of the secondary radiolysis inside fuel canister is neglected. It is also assumed that secondary phases of radionuclides do not precipitate inside the canister. The model results show that a relatively large domain of the near-field can be oxidised by the oxidants of secondary radiolysis. In the second case it is assumed that the radionuclide concentration within the canister is controlled by its respective solubility limit. The amount of radionuclides released out of the canister will then be limited by the solubility of the secondary phases. The effect of the secondary radiolysis will be quite limited in

  2. Analysis of sludge from Hanford K East Basin canisters

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J. [ed.] [comp.] [DE and S Hanford, Inc., Richland, WA (United States); Welsh, T.L. [B and W Protec, Inc. (United States); Baker, R.B. [DE and S Hanford, Inc., Richland, WA (United States); Hoppe, E.W.; Schmidt, A.J.; Abrefah, J.; Tingey, J.M.; Bredt, P.R.; Golcar, G.R. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-09-12

    Sludge samples from the canisters in the Hanford K East Basin fuel storage pool have been retrieved and analyzed. Both chemical and physical properties have been determined. The results are to be used to determine the disposition of the bulk of the sludge and to assess the impact of residual sludge on dry storage of the associated intact metallic uranium fuel elements. This report is a summary and review of the data provided by various laboratories. Although raw chemistry data were originally reported on various bases (compositions for as-settled, centrifuged, or dry sludge) this report places all of the data on a common comparable basis. Data were evaluated for internal consistency and consistency with respect to the governing sample analysis plan. Conclusions applicable to sludge disposition and spent fuel storage are drawn where possible.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Molecular Contamination on Anodized Aluminum Components of the Genesis Science Canister

    Science.gov (United States)

    Burnett, D. S.; McNamara, K. M.; Jurewicz, A.; Woolum, D.

    2005-01-01

    Inspection of the interior of the Genesis science canister after recovery in Utah, and subsequently at JSC, revealed a darkening on the aluminum canister shield and other canister components. There has been no such observation of film contamination on the collector surfaces, and preliminary spectroscopic ellipsometry measurements support the theory that the films observed on the anodized aluminum components do not appear on the collectors to any significant extent. The Genesis Science Team has made an effort to characterize the thickness and composition of the brown stain and to determine if it is associated with molecular outgassing.Detailed examination of the surfaces within the Genesis science canister reveals that the brown contamination is observed to varying degrees, but only on surfaces exposed in space to the Sun and solar wind hydrogen. In addition, the materials affected are primarily composed of anodized aluminum. A sharp line separating the sun and shaded portion of the thermal closeout panel is shown. This piece was removed from a location near the gold foil collector within the canister. Future plans include a reassembly of the canister components to look for large-scale patterns of contamination within the canister to aid in revealing the root cause.

  5. Infrared Spectroscopy of Gas-Phase M(+)(CO2)n (M = Co, Rh, Ir) Ion-Molecule Complexes.

    Science.gov (United States)

    Iskra, Andreas; Gentleman, Alexander S; Kartouzian, Aras; Kent, Michael J; Sharp, Alastair P; Mackenzie, Stuart R

    2017-01-12

    The structures of gas-phase M(+)(CO2)n (M = Co, Rh, Ir; n = 2-15) ion-molecule complexes have been investigated using a combination of infrared resonance-enhanced photodissociation (IR-REPD) spectroscopy and density functional theory. The results provide insight into fundamental metal ion-CO2 interactions, highlighting the trends with increasing ligand number and with different group 9 ions. Spectra have been recorded in the region of the CO2 asymmetric stretch around 2350 cm(-1) using the inert messenger technique and their interpretation has been aided by comparison with simulated infrared spectra of calculated low-energy isomeric structures. All vibrational bands in the smaller complexes are blue-shifted relative to the asymmetric stretch in free CO2, consistent with direct binding to the metal center dominated by charge-quadrupole interactions. For all three metal ions, a core [M(+)(CO2)2] structure is identified to which subsequent ligands are less strongly bound. No evidence is observed in this size regime for complete activation or insertion reactions.

  6. Heat flux measurements of Tb3M series (M=Co, Rh and Ru): Specific heat and magnetocaloric properties

    Science.gov (United States)

    Monteiro, J. C. B.; Lombardi, G. A.; dos Reis, R. D.; Freitas, H. E.; Cardoso, L. P.; Mansanares, A. M.; Gandra, F. G.

    2016-12-01

    We report on the magnetic properties and magnetocaloric effect (MCE) for the Tb3M series, with M=Co, Rh and Ru, obtained using a heat flux technique. The specific heat of Tb3Co and Tb3Rh are very similar, with a first order type transition occurring around 6 K below the magnetic ordering temperature without any corresponding feature on the magnetization. The slightly enhanced electronic specific heat, the Debye temperature around 150 K and the presence of the magnetic specific heat well above the ordering temperature are also characteristic of many other compounds of the R3M family (R=Rare Earth). The specific heat for Tb3Ru, however, presents two peaks at 37 K and 74 K. The magnetization shows that below the first peak the system presents an antiferromagnetic behavior and is paramagnetic above 74 K. We obtained a magnetocaloric effect for M=Co and Rh, -∆S=12 J/kg K, but for Tb3Ru it is less than 3 J/kg K (μ0∆H=5 T). We believe that the experimental results show that the MCE is directly related with the process of hybridization of the (R)5d-(M)d electrons that occurs in the R3M materials.

  7. Evaluation of the potential for gas pressurization and free liquid accumulation in a WVDP canister

    Energy Technology Data Exchange (ETDEWEB)

    Hazelton, R.F.; Thornhill, C.K.

    1993-12-01

    A full-scale canister provided by the West Valley Demonstration Project, filled during the SF-11 vitrification qualification test, was tested to determine its potential for gas generation (non-radiolitic only) and liquid accumulation. The canister was sealed and held at a temperature of about 500{degrees}C for eight weeks. Gas samples obtained during the test were analyzed using mass spectroscopy to determine the composition of gases within the canister. At the end of the eight weeks the canister gases were evacuated through a desiccant to capture any water that had been released by the glass during the test. In addition, an analysis of the glass using fourier transform infrared spectroscopy was performed to determine the water content in the glass both before and after the temperature exposure.

  8. HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, D.R.

    1999-10-07

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments.

  9. Cold Vacuum Drying (CVD) OCRWM Loop Error Determination

    Energy Technology Data Exchange (ETDEWEB)

    PHILIPP, B.L.

    2000-07-26

    Characterization is specifically identified by the Richland Operations Office (RL) for the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE), as requiring application of the requirements in the Quality Assurance Requirements and Description (QARD) (RW-0333P DOE 1997a). Those analyses that provide information that is necessary for repository acceptance require application of the QARD. The cold vacuum drying (CVD) project identified the loops that measure, display, and record multi-canister overpack (MCO) vacuum pressure and Tempered Water (TW) temperature data as providing OCRWM data per Application of the Office of Civilian Radioactive Waste Management (OCRWM) Quality Assurance Requirements to the Hanford Spent Nuclear Fuel Project HNF-SD-SNF-RPT-007. Vacuum pressure transmitters (PT 1*08, 1*10) and TW temperature transmitters (TIT-3*05, 3*12) are used to verify drying and to determine the water content within the MCO after CVD.

  10. Draft report: Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions.

  11. Physical properties of encapsulate spent fuel in canisters; Comportamiento fisico de las capsulas de almacenamiento

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

  12. Design basis for the copper canister. Stage one

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W. H. [ERA Technology Limited, Leatherhead, Surrey (United Kingdom)

    1995-02-01

    The copper/iron canister which has been proposed for containment of high level waste in the Swedish Nuclear Waste Disposal Programme has been studied from the points of view of choice of materials, manufacturing technology and quality assurance. The choice of High Strength Low Alloy steel for the load bearing element appears to be a good choice but it is necessary to understand the effect of laser welding on the structure of the chosen alloy and to ensure that the very rapid cooling rates which attend laser welding of thick material do not lead to the development of untempered martensite. The choice of an almost pure copper for the corrosion barrier is based on the very good corrosion resistance claimed for it under repository conditions. Production trials are in progress using this material and serious difficulties are expected both in manufacture and in quality assurance. The trials may or may not produce a satisfactory prototype but they will give pointers towards modifications in choice of material and processing technology. This study concludes that the chosen material is particularly difficult to process and to test, and that the claimed good corrosion resistance in in doubt. 54 refs.

  13. Inorganic analyses of volatilized and condensed species within prototypic Defense Waste Processing Facility (DWPF) canistered waste

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.M.

    1992-06-30

    The high-level radioactive waste currently stored in carbon steel tanks at the Savannah River Site (SRS) will be immobilized in a borosilicate glass in the Defense Waste Processing Facility (DWPF). The canistered waste will be sent to a geologic repository for final disposal. The Waste Acceptance Preliminary Specifications (WAPS) require the identification of any inorganic phases that may be present in the canister that may lead to internal corrosion of the canister or that could potentially adversely affect normal canister handling. During vitrification, volatilization of mixed (Na, K, Cs)Cl, (Na, K, Cs){sub 2}SO{sub 4}, (Na, K, Cs)BF{sub 4}, (Na, K){sub 2}B{sub 4}O{sub 7} and (Na,K)CrO{sub 4} species from glass melt condensed in the melter off-gas and in the cyclone separator in the canister pour spout vacuum line. A full-scale DWPF prototypic canister filled during Campaign 10 of the SRS Scale Glass Melter was sectioned and examined. Mixed (NaK)CI, (NaK){sub 2}SO{sub 4}, (NaK) borates, and a (Na,K) fluoride phase (either NaF or Na{sub 2}BF{sub 4}) were identified on the interior canister walls, neck, and shoulder above the melt pour surface. Similar deposits were found on the glass melt surface and on glass fracture surfaces. Chromates were not found. Spinel crystals were found associated with the glass pour surface. Reference amounts of the halides and sulfates were found retained in the glass and the glass chemistry, including the distribution of the halides and sulfates, was homogeneous. In all cases where rust was observed, heavy metals (Zn, Ti, Sn) from the cutting blade/fluid were present indicating that the rust was a reaction product of the cutting fluid with glass and heat sensitized canister or with carbon-steel contamination on canister interior. Only minimal water vapor is present so that internal corrosion of the canister, will not occur.

  14. Description of Defense Waste Processing Facility reference waste form and canister. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, R.G.

    1983-08-01

    The Defense Waste Processing Facility (DWPF) will be located at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1984. The reference waste form is borosilicate glass containing approx. 28 wt % sludge oxides, with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains about 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. Leachabilities of SRP waste glasses are expected to approach 10/sup -8/ g/m/sup 2/-day based upon 1000-day tests using glasses containing SRP radioactive waste. Tests were performed under a wide variety of conditions simulating repository environments. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approx. 470 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the sludge and supernate processes. The radionuclide content of the canister is about 177,000 ci, with a radiation level of 5500 rem/h at canister surface contact. The reference canister is fabricated of standard 24-in.-OD, Schedule 20, 304L stainless steel pipe with a dished bottom, domed head, and a combined lifting and welding flange on the head neck. The overall canister length is 9 ft 10 in. with a 3/8-in. wall thickness. The 3-m canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected as an optimum size from glass quality considerations, a logical size for repository handling and to ensure that a filled canister with its double containment shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be compatible with preliminary assessments of repository requirements. 10 references.

  15. System-Level Logistics for Dual Purpose Canister Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kalinina, Elena A.

    2014-06-03

    The analysis presented in this report investigated how the direct disposal of dual purpose canisters (DPCs) may be affected by the use of standard transportation aging and disposal canisters (STADs), early or late start of the repository, and the repository emplacement thermal power limits. The impacts were evaluated with regard to the availability of the DPCs for emplacement, achievable repository acceptance rates, additional storage required at an interim storage facility (ISF) and additional emplacement time compared to the corresponding repackaging scenarios, and fuel age at emplacement. The result of this analysis demonstrated that the biggest difference in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario is for a repository start date of 2036 with a 6 kW thermal power limit. The differences are also seen in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario for the alternative with a 6 kW thermal limit and a 2048 start date, and for the alternatives with a 10 kW thermal limit and 2036 and 2048 start dates. The alternatives with disposal of UNF in both DPCs and STADs did not require additional storage, regardless of the repository acceptance rate, as compared to the reference repackaging case. In comparison to the reference repackaging case, alternatives with the 18 kW emplacement thermal limit required little to no additional emplacement time, regardless of the repository start time, the fuel loading scenario, or the repository acceptance rate. Alternatives with the 10 kW emplacement thermal limit and the DPCs and STADs fuel loading scenario required some additional emplacement time. The most significant decrease in additional emplacement time occurred in the alternative with the 6 kW thermal limit and the 2036 repository starting date. The average fuel age at emplacement ranges from 46 to 88 years. The maximum fuel age at

  16. Tests for manufacturing technology of disposal canisters for nuclear spent fuel; Loppusijoituskapselin valmistustekniset kokeet

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy (Finland); Salonen, T. [Outokumpu Poricopper Oy (Finland); Meuronen, I. [Suomen Teknohaus Oy (Finland); Lehto, K. [Valmet Oyj Rautpohja Foundry (Finland)

    1999-06-01

    The summary and status of the results of the manufacturing technology programmes concerning the disposal canister for spent nuclear fuel conducted by Posiva Oy are given in this report. Posiva has maintained a draft plan for a disposal canister design and an assessment of potential manufacturing technologies for about ten years in Finland. Now, during the year 1999, the first full scale demonstration canister is manufactured in Finland. The technology used for manufacturing of this prototype is developed by Posiva Oy mainly in co-operation with domestic industry. The main partner in developing the manufacturing technology for the copper shell has been Outokumpu Poricopper Oy, Pori, Finland, and the main partner in developing the technology for the iron insert of the canister has been Valmet Oyj Rautpohja Foundry, Jyvaeskylae, Finland. In both areas many subcontractors have been used, predominantly domestic engineering workshops, but also some foreign subcontractors, e.g. for EB-welding, who have had large enough welding equipment. This report describes the developing programmes for canister manufacturing, evaluates the results and presents some alternative methods, and tries to evaluate the pros and contras of them. In addition, the adequacy of the achieved technological know-how is assessed in respect of the required quality of the disposal canister. The following manufacturing technologies have been the concrete topics of the development programme: Electron beam welding technology development for thick-walled copper, Casting of massive copper billets, Hot rolling of thick-walled copper plates, Hot pressing and forging in lid manufacture, Extrusion and drawing of copper tubes, Bending of copper plates by roller or press, Machining of copper, Residual stress removal by heat treatment, Non-destructive testing, Long-term strength of EB-welds, Casting and machining of the iron insert of the canister The specialists from all the main developing partner companies have

  17. Progress in the understanding of the long-term corrosion behaviour of copper canisters

    Science.gov (United States)

    King, Fraser; Lilja, Christina; Vähänen, Marjut

    2013-07-01

    Copper has been proposed as a canister material for the disposal of spent nuclear fuel in a deep geologic repository in a number of countries worldwide. Since it was first proposed for this purpose in 1978, a significant number of studies have been performed to assess the corrosion performance of copper under repository conditions. These studies are critically reviewed and the suitability of copper as a canister material for nuclear waste is re-assessed. Over the past 30-35 years there has been considerable progress in our understanding of the expected corrosion behaviour of copper canisters. Crucial to this progress has been the improvement in the understanding of the nature of the repository environment and how it will evolve over time. With this improved understanding, it has been possible to predict the evolution of the corrosion behaviour from the initial period of warm, aerobic conditions in the repository to the long-term phase of cool, anoxic conditions dominated by the presence of sulphide. An historical review of the treatment of the corrosion behaviour of copper canisters is presented, from the initial corrosion assessment in 1978, through a major review of the corrosion behaviour in 2001, through to the current level of understanding based on the results of on-going studies. Compared with the initial corrosion assessment, there has been considerable progress in the treatment of localised corrosion, stress corrosion cracking, and microbiologically influenced corrosion of the canisters. Progress in the mechanistic modelling of the evolution of the corrosion behaviour of the canister is also reviewed, as is the continuing debate about the thermodynamic stability of copper in pure water. The overall conclusion of this critical review is that copper is a suitable material for the disposal of spent nuclear fuel and offers the prospect of containment of the waste for an extended period of time. The corrosion behaviour is influenced by the presence of the

  18. Development of a Universal Canister for Disposal of High-Level Waste in Deep Boreholes.

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gomberg, Steve [USDOE, Washington, DC (United States)

    2015-11-01

    The mission of the United States Department of Energy’s Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research. Some of the wastes that must be managed have been identified as good candidates for disposal in a deep borehole in crystalline rock. In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister-based system that can be used for handling these wastes during the disposition process (i.e., storage, transfer, transportation, and disposal) could facilitate the eventual disposal of these wastes. Development of specifications for the universal canister system will consider the regulatory requirements that apply to storage, transportation, and disposal of the capsules, as well as operational requirements and limits that could affect the design of the canister (e.g., deep borehole diameter). In addition, there are risks and technical challenges that need to be recognized and addressed as Universal Canister system specifications are developed. This paper provides an approach to developing specifications for such a canister system that is integrated with the overall efforts of the DOE’s Used Fuel Disposition Campaign's Deep Borehole Field Test and compatible with planned storage of potential borehole-candidate wastes.

  19. A New Frangible Composite Canister Cover with the Function of Specified Direction Separation

    Science.gov (United States)

    Zhou, Guangming; Cai, Deng'an; Qian, Yuan; Deng, Jian; Wang, Xiaopei

    2016-08-01

    A lightweight and auto-separated canister cover is required for quick launching in some specific missile launchers. In this paper, a new frangible composite canister cover with the function of specified direction separation is proposed and studied via both experimental and numerical approaches. The frangible canister cover with local non-split weak zone structure, which is manufactured by traditional hand lay-up process with vacuum assisted resin infusion (VARI) method, is designed to fail and separate in a predetermined and specified direction in comparison with the cover with full split weak zone structure. This design is innovative and also necessary for reduction of potential risk to peripheral equipment around the missile launcher. The failure pressure of the cover is determined on the basis of the failure criteria used in finite element (FE) model. In experimental pressurized testing, a number of frangible canister covers subjected to pressure loadings in six cases are studied. Close agreements between the experimental and numerical results have been examined. The frangible canister covers with local non-split weak zone structure which have been studied can be separated and fly out to the specified direction.

  20. Design package test weights for fuel retrieval system (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1999-10-26

    This is a design package that documents the development of test weights used in the Spent Nuclear Fuels subproject Fuel Retrieval System. The K Basins Spent Nuclear Fuel (SNF) project consists of the safe retrieval, preparation, and repackaging of the spent fuel stored at the K East (KE) and K West (KW) Basins for interim safe storage in the Canister Storage Building (CSB). Multi-Canister Overpack (MCO) scrap baskets and fuel baskets will be loaded and weighed under water. The equipment used to weigh the loaded fuel baskets requires daily calibration checks, using test weights traceable to National Institute of Standards Testing (NIST) standards. The test weights have been designated as OCRWM related in accordance with HNF-SD-SNF-RF'T-007 (McCormack).

  1. Deep penetrating eddy current for copper canister inspection. Main results

    Energy Technology Data Exchange (ETDEWEB)

    Tadeusz Stepinski [TSonic, Uppsala (Sweden)

    2004-02-01

    The aim of this project was to optimize the detection and characterization of deep flaws (voids) in copper plates. Two types of voids were investigated and compared: circular and rectangular. The circular voids had the form of cylindrical cavities while the rectangular ones were cavities with a rectangular cross section. All probes were of the same type, transmit-receive transducers with four pick-ups connected in a double differential configuration. Comparison of the EC responses to circular and rectangular voids obtained using the MDF12 probe has shown that both types of voids can be characterized using phase and amplitude of their responses in the complex impedance plane. Phase of the response in the impedance plane appeared to be a reliable measure of void depth. Phase dependence on the void depth is linear (which agrees with the theory) and its slope is approx -37 deg/mm. Magnitude of the EC response contains information on the void size provided that the void depth is known. It has been shown that magnitude of the EC responses is correlated to the lengths of the rectangular voids and hole diameter, this is, similar lengths and diameters result in similar response magnitudes. It should be noted, however, that multi-differential MDF probes generate responses with different shapes for circular and rectangular voids. First, shapes of the MDF probe responses in the impedance plane depend on the probe's orientation with respect to scanning direction. Second, they also depend upon the direction of scanning with respect to the void orientation. The measurements presented in this report were performed for the probe axis aligned along with the scanning direction and, in case of rectangular voids, for scanning direction along the void lengths. Comparison of the responses obtained from flat bottom holes in copper material taken from different canister parts has not shown any essential differences between the material samples. Conductivity measurement performed using

  2. Corrosion of the copper canister in the repository environment

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P.; Eriksson, Sture [Studsvik Material AB, Nykoeping (Sweden)

    1999-12-01

    The present report accounts for studies on copper corrosion performed at Studsvik Material AB during 1997-1999 on commission by SKI. The work has been focused on localised corrosion and electrochemistry of copper in the repository environment. The current theory of localised copper corrosion is not consistent with recent practical experiences. It is therefore desired to complete and develop the theory based on knowledge about the repository environment and evaluations of previous as well as recent experimental and field results. The work has therefore comprised a thorough compilation and up-date of literature on copper corrosion and on the repository environment. A selection of a 'working environment', defining the chemical parameters and their ranges of variation has been made and is used as a fundament for the experimental part of the work. Experiments have then been performed on the long-range electrochemical behaviour of copper in selected environments simulating the repository. Another part of the work has been to further develop knowledge about the thermodynamic limits for corrosion in the repository environment. Some of the thermodynamic work is integrated here. Especially thermodynamics for the system Cu-Cl-H-O up to 150 deg C and high chloride concentrations are outlined. However, there is also a rough overview of the whole system Cu-Fe-Cl-S-C-H-O as a fundament for the discussion. Data are normally accounted as Pourbaix diagrams. Some of the conclusions are that general corrosion on copper will probably not be of significant importance in the repository as far as transportation rates are low. However, if such rates were high, general corrosion could be disastrous, as there is no passivation of copper in the highly saline environment. The claim on knowledge of different kinds of localised corrosion and pitting is high, as pitting damages can shorten the lifetime of a canister dramatically. Normal pitting can happen in oxidising environment, but

  3. Three-Dimensional Heat Transfer Analysis for A Thermal Energy Storage Canister

    Institute of Scientific and Technical Information of China (English)

    Hou Xinbin; Xin Yuming; Yang Chunxin; Yuan Xiugan; Dong Keyong

    2001-01-01

    High temperature latent thermal storage is one of the critical techniques for a solar dynamic power system. This paper presents results from heat transfer analysis of a phase change salt containment canister. A three dimensional analysis program is developed to model heat transfer in a PCM canister. Analysis include effects of asymmetric circumference heat flux, conduction in canister walls, liquid PCM and solid PCM, void volume change and void location, and conduction and radiation across PCM vapor void. The PCM phase change process is modeled using the enthalpy method and the simulation results are compared with those of other two dimensional investigations. It's shown that there are large difference with two-dimensional analysis, therefore the three-dimensional model is necessary for system design of high temperature latent thermal storage.

  4. Uncertainty quantification methodologies development for stress corrosion cracking of canister welds

    Energy Technology Data Exchange (ETDEWEB)

    Dingreville, Remi Philippe Michel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-09-30

    This letter report presents a probabilistic performance assessment model to evaluate the probability of canister failure (through-wall penetration) by SCC. The model first assesses whether environmental conditions for SCC – the presence of an aqueous film – are present at canister weld locations (where tensile stresses are likely to occur) on the canister surface. Geometry-specific storage system thermal models and weather data sets representative of U.S. spent nuclear fuel (SNF) storage sites are implemented to evaluate location-specific canister surface temperature and relative humidity (RH). As the canister cools and aqueous conditions become possible, the occurrence of corrosion is evaluated. Corrosion is modeled as a two-step process: first, pitting is initiated, and the extent and depth of pitting is a function of the chloride surface load and the environmental conditions (temperature and RH). Second, as corrosion penetration increases, the pit eventually transitions to a SCC crack, with crack initiation becoming more likely with increasing pit depth. Once pits convert to cracks, a crack growth model is implemented. The SCC growth model includes rate dependencies on both temperature and crack tip stress intensity factor, and crack growth only occurs in time steps when aqueous conditions are predicted. The model suggests that SCC is likely to occur over potential SNF interim storage intervals; however, this result is based on many modeling assumptions. Sensitivity analyses provide information on the model assumptions and parameter values that have the greatest impact on predicted storage canister performance, and provide guidance for further research to reduce uncertainties.

  5. Radon-222 activity flux measurement using activated charcoal canisters: revisiting the methodology.

    Science.gov (United States)

    Alharbi, Sami H; Akber, Riaz A

    2014-03-01

    The measurement of radon ((222)Rn) activity flux using activated charcoal canisters was examined to investigate the distribution of the adsorbed (222)Rn in the charcoal bed and the relationship between (222)Rn activity flux and exposure time. The activity flux of (222)Rn from five sources of varying strengths was measured for exposure times of one, two, three, five, seven, 10, and 14 days. The distribution of the adsorbed (222)Rn in the charcoal bed was obtained by dividing the bed into six layers and counting each layer separately after the exposure. (222)Rn activity decreased in the layers that were away from the exposed surface. Nevertheless, the results demonstrated that only a small correction might be required in the actual application of charcoal canisters for activity flux measurement, where calibration standards were often prepared by the uniform mixing of radium ((226)Ra) in the matrix. This was because the diffusion of (222)Rn in the charcoal bed and the detection efficiency as a function of the charcoal depth tended to counterbalance each other. The influence of exposure time on the measured (222)Rn activity flux was observed in two situations of the canister exposure layout: (a) canister sealed to an open bed of the material and (b) canister sealed over a jar containing the material. The measured (222)Rn activity flux decreased as the exposure time increased. The change in the former situation was significant with an exponential decrease as the exposure time increased. In the latter case, lesser reduction was noticed in the observed activity flux with respect to exposure time. This reduction might have been related to certain factors, such as absorption site saturation or the back diffusion of (222)Rn gas occurring at the canister-soil interface.

  6. Coupled Transport/Reaction Modelling of Copper Canister Corrosion Aided by Microbial Processes

    Energy Technology Data Exchange (ETDEWEB)

    Jinsong Liu [Royal Institute of Technology, Stockholm (Sweden). Dept. of Chemical Engineering and Technology

    2006-04-15

    Copper canister corrosion is an important issue in the concept of a nuclear fuel repository. Previous studies indicate that the oxygen-free copper canister could hold its integrity for more than 100,000 years in the repository environment. Microbial processes may reduce sulphate to sulphide and considerably increase the amount of sulphides available for corrosion. In this paper, a coupled transport/reaction model is developed to account for the transport of chemical species produced by microbial processes. The corroding agents like sulphide would come not only from the groundwater flowing in a fracture that intersects the canister, but also from the reduction of sulphate near the canister. The reaction of sulphate-reducing bacteria and the transport of sulphide in the bentonite buffer are included in the model. The depth of copper canister corrosion is calculated by the model. With representative 'central values' of the concentrations of sulphate and methane at repository depth at different sites in Fennoscandian Shield the corrosion depth predicted by the model is a few millimetres during 10{sup 5} years. As the concentrations of sulphate and methane are extremely site-specific and future climate changes may significantly influence the groundwater compositions at potential repository sites, sensitivity analyses have been conducted. With a broad perspective of the measured concentrations at different sites in Sweden and in Finland, and some possible mechanisms (like the glacial meltwater intrusion and interglacial seawater intrusion) that may introduce more sulphate into the groundwater at intermediate depths during future climate changes, higher concentrations of either/both sulphate and methane than what is used as the representative 'central' values would be possible. In worst cases. locally, half of the canister thickness could possibly be corroded within 10{sup 5} years.

  7. System Configuration Management Implementation Procedure for the Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    GARRISON, R.C.

    2000-11-28

    This document provides configuration management for the Distributed Control System (DCS), the Gaseous Effluent Monitoring System (GEMS-100) System, the Heating Ventilation and Air Conditioning (HVAC) Programmable Logic Controller (PLC), the Canister Receiving Crane (CRC) CRN-001 PLC, and both North and South vestibule door interlock system PLCs at the Canister Storage Building (CSB). This procedure identifies and defines software configuration items in the CSB control and monitoring systems, and defines configuration control throughout the system life cycle. Components of this control include: configuration status accounting; physical protection and control; and verification of the completeness and correctness of these items.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Friction stir welding - an alternative method for sealing nuclear waste storage canisters

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, R.E. [TWI Ltd, Cambridge (United Kingdom)

    2004-12-01

    When welding 50 mm thick copper a very high heat input is required to combat the high thermal diffusivity and only the Electron Beam Welding (EBW) process had this capability when this copper canister concept was conceived. Despite the encouraging results achieved using EBW with thick section copper, SKB felt that it would be prudent to assess other joining methods. This assessment concluded that friction welding, could also provide very high quality welds to satisfy the service life requirements of the SKB canister design. A friction welding variant called Friction Stir Welding (FSW) was shown to have the capability of welding 3 mm thick copper sheet with excellent integrity and reproducibility. This later provided sufficient encouragement for SKB to consider the potential of FSW as a method for joining thick section copper, using relatively simple machine tool based technology. It was thought that FSW might provide an alternative or complementary method for welding lids, or bases to canisters. In 1997 an FSW development programme started at TWI, focussed on the feasibility of welding 10 mm thick copper plate. Once this task was successfully completed, work continued to demonstrate that progressively thicker plate, up to 50 mm thick, could be joined. At this stage, with process viability established, a full size experimental FSW canister machine was designed and built. Work with this machine finished in January 2003, when it had been shown that FSW could definitely be used to weld lids to full size canisters. This report summarises the TWI development of FSW for SKB from 1997 to January 2003. It also highlights the important aspects of the process and the project milestones that will help to ensure that SKB has a welding technology that can be used with confidence for production fabrication of copper waste storage canisters in the future. The overall conclusion to this FSW development is that there is no doubt that the FSW process could be used to produce full

  10. Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-12-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions. To evaluate the potential environment on the surface of the canisters, SNL is working with the Electric Power Research Institute (EPRI) to collect and analyze dust samples from the surface of in-service SNF storage canisters. In FY 13, SNL analyzed samples from the Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI); here, results are presented for samples collected from two additional near-marine ISFSI sites, Hope Creek NJ, and Diablo Canyon CA. The Hope Creek site is located on the shores of the Delaware River within the tidal zone; the water is brackish and wave action is normally minor. The Diablo Canyon site is located on a rocky Pacific Ocean shoreline with breaking waves. Two types of samples were collected: SaltSmart™ samples, which leach the soluble salts from a known surface area of the canister, and dry pad samples, which collected a surface salt and dust using a swipe method with a mildly abrasive ScotchBrite™ pad. The dry samples were used to characterize the mineralogy and texture of the soluble and insoluble components in the dust via microanalytical techniques, including mapping X-ray Fluorescence spectroscopy and Scanning Electron Microscopy. For both Hope Creek and Diablo Canyon canisters, dust loadings were much higher on the flat upper surfaces of the canisters than on the vertical sides. Maximum dust sizes collected at both sites were slightly larger than 20 μm, but Phragmites grass seeds ~1 mm in size, were observed on the tops of the Hope Creek canisters

  11. Deposition of LaMO 3 (M=Co, Cr, Al) films by spray pyrolysis in inductively coupled plasma

    Science.gov (United States)

    Ichinose, Hiromichi; Katsuki, Hiroaki; Nagano, Masamitsu

    1994-11-01

    LaMO 3 (M=Co, Cr, Al) films were prepared on substrates by introducing ultrasonically atomized metal nitrate solutions into an inductively coupled plasma under atmospheric pressure (spray-ICP technique). Dense perovskite-type oxide films of LaCoO 3 and LaCrO 3 were obtained at 600-900°C, while the LaAiO 3 films consisted of loosely packed aggregates. Deposition rates of the films were 6-35 nm/min at 600-900°C. The high temperature phases (cubic) of LaCoO 3 and LaAlO 3 crystallized due to effect of grain size. LaCrO 3 film crystallized in the room temperature phase (orthorhombic). LaCoO 3 was highly oriented to (100) on MgO(100), and LaCrO 3 to (011) and (101) on sapphire(001). Lowest electric resistivities of LaCoO 3 and LaCrO 3 film on MgO were 9.8X10 -3 and 2.7X10 -1 Ω m, respectively, at room temperature.

  12. Quantitative ion beam analysis of M-C-O systems: application to an oxidized uranium carbide sample

    Science.gov (United States)

    Martin, G.; Raveu, G.; Garcia, P.; Carlot, G.; Khodja, H.; Vickridge, I.; Barthe, M. F.; Sauvage, T.

    2014-04-01

    A large variety of materials contain both carbon and oxygen atoms, in particular oxidized carbides, carbon alloys (as ZrC, UC, steels, etc.), and oxycarbide compounds (SiCO glasses, TiCO, etc.). Here a new ion beam analysis methodology is described which enables quantification of elemental composition and oxygen concentration profile over a few microns. It is based on two procedures. The first, relative to the experimental configuration relies on a specific detection setup which is original in that it enables the separation of the carbon and oxygen NRA signals. The second concerns the data analysis procedure i.e. the method for deriving the elemental composition from the particle energy spectrum. It is a generic algorithm and is here successfully applied to characterize an oxidized uranium carbide sample, developed as a potential fuel for generation IV nuclear reactors. Furthermore, a micro-beam was used to simultaneously determine the local elemental composition and oxygen concentration profiles over the first microns below the sample surface. This method is adapted to the determination of the composition of M?C?O? compounds with a sensitivity on elemental atomic concentrations around 1000 ppm.

  13. Multi-dimensional modeling of a thermal energy storage canister. M.S. Thesis - Cleveland State Univ., Dec. 1990

    Science.gov (United States)

    Kerslake, Thomas W.

    1991-01-01

    The Solar Dynamic Power Module being developed for Space Station Freedom uses a eutectic mixture of LiF-CaF2 phase change material (PCM) contained in toroidal canisters for thermal energy storage. Presented are the results from heat transfer analyses of a PCM containment canister. One and two dimensional finite difference computer models are developed to analyze heat transfer in the canister walls, PCM, void, and heat engine working fluid coolant. The modes of heat transfer considered include conduction in canister walls and solid PCM, conduction and pseudo-free convection in liquid PCM, conduction and radiation across PCM vapor filled void regions, and forced convection in the heat engine working fluid. Void shape, location, growth or shrinkage (due to density difference between the solid and liquid PCM phases) are prescribed based on engineering judgment. The PCM phase change process is analyzed using the enthalpy method. The discussion of the results focuses on how canister thermal performance is affected by free convection in the liquid PCM and void heat transfer. Characterizing these effects is important for interpreting the relationship between ground-based canister performance (in 1-g) and expected on-orbit performance (in micro-g). Void regions accentuate canister hot spots and temperature gradients due to their large thermal resistance. Free convection reduces the extent of PCM superheating and lowers canister temperatures during a portion of the PCM thermal charge period. Surprisingly small differences in canister thermal performance result from operation on the ground and operation on-orbit. This lack of a strong gravity dependency is attributed to the large contribution of container walls in overall canister energy redistribution by conduction.

  14. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  15. Instrumentation. Nondestructive Examination for Verification of Canister and Cladding Integrity - FY2013 Status Update

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Anthony M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Denslow, Kayte M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crawford, Susan L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This report documents FY13 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) characterize hydrogen effects in UNF cladding to facilitate safe storage and retrieval.

  16. Instrumentation: Nondestructive Examination for Verification of Canister and Cladding Integrity. FY2014 Status Update

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suter, Jonathan D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Anthony M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-12

    This report documents FY14 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) verify the integrity of dry storage cask internals.

  17. Fuel and canister process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars; Lilja, Christina (eds.)

    2010-12-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  18. Miniature Canister (MiniCan) Corrosion experiment progress report 4 for 2008-2011

    Energy Technology Data Exchange (ETDEWEB)

    Smart, Nick; Reddy, Bharti; Rance, Andy [Serco, Hook (United Kingdom)

    2012-06-15

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB of Sweden are considering using the Copper-Iron Canister, which consists of an outer copper canister and a cast iron insert. Over the years a programme of laboratory work has been carried out to investigate a range of corrosion issues associated with the canister, including the possibility of expansion of the outer copper canister as a result of the anaerobic corrosion of the cast iron insert. Previous experimental work using stacks of test specimens has not shown any evidence of corrosion-induced expansion. However, as a further step in developing an understanding of the likely performance of the canister in a repository environment, Serco has set up a series of experiments in SKB's Aespoe Hard Rock Laboratory (HRL) using inactive model canisters, in which leaks were deliberately introduced into the outer copper canister while surrounded by bentonite, with the aim of obtaining information about the internal corrosion evolution of the internal environment. The experiments use five small scale model canisters (300 mm long x 150 mm diameter) that simulate the main features of the SKB canister design (hence the project name, 'MiniCan'). The main aim of the work is to examine how corrosion of the cast iron insert will evolve if a leak is present in the outer copper canister. This report describes the progress on the five experiments running at the Aespoe Hard Rock Laboratory and the data obtained from the start of the experiments in late 2006 up to Winter 2011. The full details of the design and installation of the experiments are given in a previous report and this report concentrates on summarising and interpreting the data obtained to date. This report follows the earlier progress reports presenting results up to December 2010. The current document (progress report 4) describes work up to December 2011. The current report presents the results of the water analyses

  19. Miniature Canister (MiniCan) Corrosion Experiment Progress Report 3 for 2008-2010

    Energy Technology Data Exchange (ETDEWEB)

    Smart, N.R.; Reddy, B.; Rance, A.P. (Serco (United Kingdom))

    2011-08-15

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB of Sweden are considering using the Copper-Iron Canister, which consists of an outer copper canister and a cast iron insert. Over the years a programme of laboratory work has been carried out to investigate a range of corrosion issues associated with the canister, including the possibility of expansion of the outer copper canister as a result of the anaerobic corrosion of the cast iron insert. Previous experimental work using stacks of test specimens has not shown any evidence of corrosion-induced expansion. However, as a further step in developing an understanding of the likely performance of the canister in a repository environment, Serco has set up a series of experiments in SKB's Aespoe Hard Rock Laboratory (HRL) using inactive model canisters, in which leaks were deliberately introduced into the outer copper canister while surrounded by bentonite, with the aim of obtaining information about the internal corrosion evolution of the internal environment. The experiments use five small-scale model canisters (300 mm long x 150 mm diameter) that simulate the main features of the SKB canister design (hence the project name, 'MiniCan'). The main aim of the work is to examine how corrosion of the cast iron insert will evolve if a leak is present in the outer copper canister. This report describes the progress on the five experiments running at the Aespoe Hard Rock Laboratory and the data obtained from the start of the experiments in late 2006 up to Winter 2010. The full details of the design and installation of the experiments are given in a previous report and this report concentrates on summarising and interpreting the data obtained to date. This report follows two earlier progress reports presenting results up to December 2009. The current document (progress report 3) describes work up to December 2010. The current report presents the results of the water analyses

  20. Clean Assembly of Genesis Collector Canister for Flight: Lessons for Planetary Sample Return

    Science.gov (United States)

    Allton, J. H.; Stansbery, E. K.; Allen, C. C.; Warren, J. L.; Schwartz, C. M.

    2007-01-01

    Measurement of solar composition in the Genesis collectors requires not only high sensitivity but very low blanks; thus, very strict collector contamination minimization was required beginning with mission planning and continuing through hardware design, fabrication, assembly and testing. Genesis started with clean collectors and kept them clean inside of a canister. The mounting hardware and container for the clean collectors were designed to be cleanable, with access to all surfaces for cleaning. Major structural components were made of aluminum and cleaned with megasonically energized ultrapure water (UPW). The UPW purity was >18 M resistivity. Although aluminum is relatively difficult to clean, the Genesis protocol achieved level 25 and level 50 cleanliness on large structural parts; however, the experience suggests that surface treatments may be helpful on future missions. All cleaning was performed in an ISO Class 4 (Class 10) cleanroom immediately adjacent to an ISO Class 4 assembly room; thus, no plastic packaging was required for transport. Persons assembling the canister were totally enclosed in cleanroom suits with face shield and HEPA filter exhaust from suit. Interior canister materials, including fasteners, were installed, untouched by gloves, using tweezers and other stainless steel tools. Sealants/lubricants were not exposed inside the canister, but vented to the exterior and applied in extremely small amounts using special tools. The canister was closed in ISO Class 4, not to be opened until on station at Earth-Sun L1. Throughout the cleaning and assembly, coupons of reference materials that were cleaned at the same time as the flight hardware were archived for future reference and blanks. Likewise reference collectors were archived. Post-mission analysis of collectors has made use of these archived reference materials.

  1. Evaluation of DUSTRAN Software System for Modeling Chloride Deposition on Steel Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Tracy T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fritz, Brad G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rutz, Frederick C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Devanathan, Ram [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-29

    The degradation of steel by stress corrosion cracking (SCC) when exposed to atmospheric conditions for decades is a significant challenge in the fossil fuel and nuclear industries. SCC can occur when corrosive contaminants such as chlorides are deposited on a susceptible material in a tensile stress state. The Nuclear Regulatory Commission has identified chloride-induced SCC as a potential cause for concern in stainless steel used nuclear fuel (UNF) canisters in dry storage. The modeling of contaminant deposition is the first step in predictive multiscale modeling of SCC that is essential to develop mitigation strategies, prioritize inspection, and ensure the integrity and performance of canisters, pipelines, and structural materials. A multiscale simulation approach can be developed to determine the likelihood that a canister would undergo SCC in a certain period of time. This study investigates the potential of DUSTRAN, a dust dispersion modeling system developed by Pacific Northwest National Laboratory, to model the deposition of chloride contaminants from sea salt aerosols on a steel canister. Results from DUSTRAN simulations run with historical meteorological data were compared against measured chloride data at a coastal site in Maine. DUSTRAN’s CALPUFF model tended to simulate concentrations higher than those measured; however, the closest estimations were within the same order of magnitude as the measured values. The decrease in discrepancies between measured and simulated values as the level of abstraction in wind speed decreased suggest that the model is very sensitive to wind speed. However, the influence of other parameters such as the distinction between open-ocean and surf-zone sources needs to be explored further. Deposition values predicted by the DUSTRAN system were not in agreement with concentration values and suggest that the deposition calculations may not fully represent physical processes. Overall, results indicate that with parameter

  2. Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-12-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions. To evaluate the potential environment on the surface of the canisters, SNL is working with the Electric Power Research Institute (EPRI) to collect and analyze dust samples from the surface of in-service SNF storage canisters. In FY 13, SNL analyzed samples from the Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI); here, results are presented for samples collected from two additional near-marine ISFSI sites, Hope Creek NJ, and Diablo Canyon CA. The Hope Creek site is located on the shores of the Delaware River within the tidal zone; the water is brackish and wave action is normally minor. The Diablo Canyon site is located on a rocky Pacific Ocean shoreline with breaking waves. Two types of samples were collected: SaltSmart™ samples, which leach the soluble salts from a known surface area of the canister, and dry pad samples, which collected a surface salt and dust using a swipe method with a mildly abrasive ScotchBrite™ pad. The dry samples were used to characterize the mineralogy and texture of the soluble and insoluble components in the dust via microanalytical techniques, including mapping X-ray Fluorescence spectroscopy and Scanning Electron Microscopy. For both Hope Creek and Diablo Canyon canisters, dust loadings were much higher on the flat upper surfaces of the canisters than on the vertical sides. Maximum dust sizes collected at both sites were slightly larger than 20 μm, but Phragmites grass seeds ~1 mm in size, were observed on the tops of the Hope Creek canisters

  3. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  4. The effect of discontinuities on the corrosion behaviour of copper canisters

    Energy Technology Data Exchange (ETDEWEB)

    King, F. [Integrity Corrosion Consulting Ltd, Calgary, AL (Canada)

    2004-03-01

    Discontinuities may remain in the weld region of copper canisters following the final closure welding and inspection procedures. Although the shell of the copper canister is expected to exhibit excellent corrosion properties in the repository environment, the question remains what impact these discontinuities might have on the long-term performance and service life of the canister. A review of the relevant corrosion literature has been carried out and an expert opinion of the impact of these discontinuities on the canister lifetime has been developed. Since the amount of oxidant in the repository is limited and the maximum wall penetration is expected to be < 2 mm, discontinuities will only be significant if they impact the localised corrosion or stress corrosion cracking (SCC) behaviour of the canister. Not all of the discontinuities will impact the corrosion behaviour of the canister. Only surface-breaking discontinuities and those discontinuities within 2 mm of the surface will affect the corrosion behaviour. Defects located further away from the finished surface will have no impact. The relevant literature on the initiation and propagation of localised corrosion and SCC has been reviewed. Initiation of localised corrosion occurs at the microscopic scale at grain boundaries, and will not be affected by the presence of macroscopic discontinuities. The localised breakdown of a passive Cu{sub 2}O/Cu(OH){sub 2} film at a critical electrochemical potential determines where and when pits initiate, not the presence of pit-shaped surface discontinuities. The factors controlling pit growth and death are well understood. There is evidence for a maximum pit radius for copper in chloride solutions, above which the small anodic: cathodic surface area ratio required for the formation of deep pits cannot be sustained. This maximum pit radius is of the order of 0.1-0.5 mm. Surface discontinuities larger than this size are unlikely to propagate as pits, and pits generated from

  5. A study of defects which might arise in the copper steel canister

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom)

    1999-05-15

    A study has been conducted to identify the material and manufacturing defects that might occur in serially produced canisters to the SKB reference design. The study has depended on cooperation of contractors engaged by SKB to participate in the development program, SKB staff, observations made by the writer over a five-year involvement with SKI, literature studies and consultation with experts. The candidate manufacturing procedures have been described inasmuch as it has been necessary to do so to make the points related to defects. Where possible, the cause of defects, their likely effects on manufacturing procedures or on durability of the canister and the methods available for their detection are given. For ease of reference each section of the report contains a table which summarizes the information in it and, in the final section of the report, all the tables are presented en-bloc.

  6. Oxidative Dissolution of Spent Fuel and Release of Nuclides from a Copper/Iron Canister : Model Developments and Applications

    OpenAIRE

    Liu, Longcheng

    2001-01-01

    Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechani...

  7. NDT Reliability - Final Report. Reliability in non-destructive testing (NDT) of the canister components

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, Mato; Takahashi, Kazunori; Mueller, Christina; Boehm, Rainer (BAM, Federal Inst. for Materials Research and Testing, Berlin (Germany)); Ronneteg, Ulf (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden))

    2008-12-15

    This report describes the methodology of the reliability investigation performed on the ultrasonic phased array NDT system, developed by SKB in collaboration with Posiva, for inspection of the canisters for permanent storage of nuclear spent fuel. The canister is composed of a cast iron insert surrounded by a copper shell. The shell is composed of the tube and the lid/base which are welded to the tube after the fuel has been place, in the tube. The manufacturing process of the canister parts and the welding process are described. Possible defects, which might arise in the canister components during the manufacturing or in the weld during the welding, are identified. The number of real defects in manufactured components have been limited. Therefore the reliability of the NDT system has been determined using a number of test objects with artificial defects. The reliability analysis is based on the signal response analysis. The conventional signal response analysis is adopted and further developed before applied on the modern ultrasonic phased-array NDT system. The concept of multi-parameter a, where the response of the NDT system is dependent on more than just one parameter, is introduced. The weakness of use of the peak signal response in the analysis is demonstrated and integration of the amplitudes in the C-scan is proposed as an alternative. The calculation of the volume POD, when the part is inspected with more configurations, is also presented. The reliability analysis is supported by the ultrasonic simulation based on the point source synthesis method

  8. ALPHN: A computer program for calculating ([alpha], n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ([alpha], n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ([alpha], n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ([alpha], n) neutron production of each actinide in neutrons per second and the total for the canister. The ([alpha], n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  9. Topical safety analysis report for the transportation of the NUHOMS{reg_sign} dry shielded canister. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS{reg_sign}) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS{reg_sign} DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS{reg_sign} Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport.

  10. Criticality Analysis for Proposed Maximum Fuel Loading in a Standardized SNF Canister with Type 1a Baskets

    Energy Technology Data Exchange (ETDEWEB)

    Chad Pope; Larry L. Taylor; Soon Sam Kim

    2007-02-01

    This document represents a summary version of the criticality analysis done to support loading SNF in a Type 1a basket/standard canister combination. Specifically, this engineering design file (EDF) captures the information pertinent to the intact condition of four fuel types with different fissile loads and their calculated reactivities. These fuels are then degraded into various configurations inside a canister without the presence of significant moderation. The important aspect of this study is the portrayal of the fuel degradation and its effect on the reactivity of a single canister given the supposition there will be continued moderation exclusion from the canister. Subsequent analyses also investigate the most reactive ‘dry’ canister in a nine canister array inside a hypothetical transport cask, both dry and partial to complete flooding inside the transport cask. The analyses also includes a comparison of the most reactive configuration to other benchmarked fuels using a software package called TSUNAMI, which is part of the SCALE 5.0 suite of software.

  11. ALPHN: A computer program for calculating ({alpha}, n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ({alpha}, n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ({alpha}, n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ({alpha}, n) neutron production of each actinide in neutrons per second and the total for the canister. The ({alpha}, n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  12. Estimates of durability of TMI-2 core debris canisters and cask liners

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Lund, A.L.; Pednekar, S.P.

    1994-04-01

    Core debris from the Three Mile Island-2 (TMI-2) reactor is currently stored in stainless steel canisters. The need to maintain the integrity of the TMI-2 core debris containers through the period of extended storage and possibly into disposal prompted this assessment. In the assessment, corrosion-induced degradation was estimated for two materials: type 304L stainless steel (SS) canisters that contain the core debris, and type 1020 carbon steel (CS) liners in the concrete casks planned for containing the canisters from 2000 AD until the TMI-2 core debris is placed in a repository. Three environments were considered: air-saturated water (with 2 ppM Cl{sup {minus}}) at 20{degree}C, and air at 20{degree}C with two relative humidities (RHs), 10 and 40%. Corrosion mechanisms assessed included general corrosion (failure criterion: 50% loss of wall thickness) and localized attack (failure criterion: through-wall pinhole penetration). Estimation of carbon steel corrosion after 50 y also was requested.

  13. Development of flaw acceptance criteria for aging management of spent nuclear fuel multi-purpose canisters

    Energy Technology Data Exchange (ETDEWEB)

    Lam, Poh -Sang [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology; Sindelar, Robert L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic in-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  14. Development of flaw acceptance criteria for aging management of spent nuclear fuel multiple-purpose canisters

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology; Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic In-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  15. New rock salt-related oxides Li{sub 3}M{sub 2}RuO{sub 6} (M=Co, Ni): Synthesis, structure, magnetism and electrochemistry

    Energy Technology Data Exchange (ETDEWEB)

    Laha, S. [Departamento de Químicas Inorganica, Facultad de Ciencias Químicas, Universidad Complutense de Madrid, 28040 Madrid (Spain); Solid State and Structural Chemistry Unit, Indian Institute of Science, Bangalore 560 012 (India); Morán, E., E-mail: emoran@quim.ucm.es [Departamento de Químicas Inorganica, Facultad de Ciencias Químicas, Universidad Complutense de Madrid, 28040 Madrid (Spain); Sáez-Puche, R.; Alario-Franco, M.Á.; Dos santos-Garcia, A.J. [Departamento de Químicas Inorganica, Facultad de Ciencias Químicas, Universidad Complutense de Madrid, 28040 Madrid (Spain); Gonzalo, E.; Kuhn, A.; García-Alvarado, F. [Universidad CEU San Pablo, Facultad de Farmacia, Departamento de Química, 28668 Boadilla del Monte, Madrid (Spain); Sivakumar, T.; Tamilarasan, S.; Natarajan, S.; Gopalakrishnan, J. [Solid State and Structural Chemistry Unit, Indian Institute of Science, Bangalore 560 012 (India)

    2013-07-15

    We describe the synthesis, crystal structure, magnetic and electrochemical characterization of new rock salt-related oxides of formula, Li{sub 3}M{sub 2}RuO{sub 6} (M=Co, Ni). The M=Co oxide adopts the LiCoO{sub 2} (R-3m) structure, where sheets of LiO{sub 6} and (Co{sub 2}/Ru)O{sub 6} octahedra are alternately stacked along the c-direction. The M=Ni oxide also adopts a similar layered structure related to Li{sub 2}TiO{sub 3}, where partial mixing of Li and Ni/Ru atoms lowers the symmetry to monoclinic (C2/c). Magnetic susceptibility measurements reveal that in Li{sub 3}Co{sub 2}RuO{sub 6}, the oxidation states of transition metal ions are Co{sup 3+} (S=0), Co{sup 2+} (S=1/2) and Ru{sup 4+} (S=1), all of them in low-spin configuration and at 10 K, the material orders antiferromagnetically. Analogous Li{sub 3}Ni{sub 2}RuO{sub 6} presents a ferrimagnetic behavior with a Curie temperature of 100 K. The differences in the magnetic behavior have been explained in terms of differences in the crystal structure. Electrochemical studies correlate well with both magnetic properties and crystal structure. Li-transition metal intermixing may be at the origin of the more impeded oxidation of Li{sub 3}Ni{sub 2}RuO{sub 6} when compared to Li{sub 3}Co{sub 2}RuO{sub 6}. Interestingly high first charge capacities (between ca. 160 and 180 mAh g{sup −1}) corresponding to ca. 2/3 of theoretical capacity are reached albeit, in both cases, capacity retention and cyclability are not satisfactory enough to consider these materials as alternatives to LiCoO{sub 2}. - Graphical abstract: Two new rock salt related oxides of formula, Li{sub 3}M{sub 2}RuO{sub 6}, (M=Co, Ni) have been prepared. The M=Co oxide adopts the LiCoO{sub 2} (R-3m) structure and the M=Ni oxide adopts a similar layered structure related to Li{sub 2}TiO{sub 3,} monoclinic (C2/c), with partial mixing of Li and Ni/Ru atoms. For Li{sub 3}Co{sub 2}RuO{sub 6}, oxidation state for Ru is 4+ and antiferromagnetic (AFM) order is

  16. Development and Application of an Oversize Reusable DOT 7A Type A Overpack Container at the Y-12 National Security Complex - 13150

    Energy Technology Data Exchange (ETDEWEB)

    Tharp, Tim [B and W Technical Services Y-12, LLC, Oak Ridge, TN 37831 (United States); Martin, David [Container Technologies Industries, LLC, Helenwood, TN 37755 (United States); Franco, Paul [Navarro Research and Engineering, Inc., Oak Ridge, TN 37831 (United States)

    2013-07-01

    Waste Management personnel at the Y-12 National Security Complex (Y-12) are concluding a multi-year effort to dispose of a large backlog of low-level waste. Six containers presented a particularly difficult technical challenge in that they each contained large robust equipment (mostly salt baths) with elevated levels of highly enriched uranium (exceeding U.S. Department of Transportation (DOT) fissile-excepted quantities). The equipment was larger than the standard 1.2 m x 1.2 m x 1.8 m (4 ft x 4 ft x 6 ft) DOT Specification 7A Type A box and would have been very difficult to size-reduce because of several inches of steel plate (along with insulating block and concrete) in the equipment design. A critical breakthrough for the success of the project involved procuring and developing two oversize reusable DOT Specification 7A Type A (fissile tested) containers (referred to as the CTI Model 7AF-690-SC) that could be used as overpacks for the original boxes of equipment. The 7A Type A overpack containers are approximately 3.5 m long x 2.7 m wide x 2.8 m high (11.7 ft x 8.9 ft x 9.2 ft) with a maximum gross weight of 10,660 kg (23,500 lb) and a payload capacity of 6,804 kg (15,000 lbs). The boxes were designed and fabricated using a split cavity design that allowed the gasketed and bolted closure to lie along the horizontal centerline of the box. The central closure location in this design allows for strengthening of box corners that tend to be points of weakness or failure in 49CFR173.465 drop tests. By combining the split cavity design with large diameter tubing and diagonal cross bracing, drop test requirements of 49CFR173.465(1) and (2) were met and demonstrated through finite element analysis modeling. The development and use of this new container dramatically reduced the need for down-sizing the equipment and allowed the project to meet objectives within cost and schedule targets. (authors)

  17. Calculation of displacements on fractures intersecting canisters induced by earthquakes: Aberg, Beberg and Ceberg examples

    Energy Technology Data Exchange (ETDEWEB)

    LaPointe, P.R.; Cladouhos, T. [Golder Associates Inc. (Sweden); Follin, S. [Golder Grundteknik KB (Sweden)

    1999-01-01

    This study shows how the method developed in La Pointe and others can be applied to assess the safety of canisters due to secondary slippage of fractures intersecting those canisters in the event of an earthquake. The method is applied to the three generic sites Aberg, Beberg and Ceberg. Estimation of secondary slippage or displacement is a four-stage process. The first stage is the analysis of lineament trace data in order to quantify the scaling properties of the fractures. This is necessary to insure that all scales of fracturing are properly represented in the numerical simulations. The second stage consists of creating stochastic discrete fracture network (DFN) models for jointing and small faulting at each of the generic sites. The third stage is to combine the stochastic DFN model with mapped lineament data at larger scales into data sets for the displacement calculations. The final stage is to carry out the displacement calculations for all of the earthquakes that might occur during the next 100,000 years. Large earthquakes are located along any lineaments in the vicinity of the site that are of sufficient size to accommodate an earthquake of the specified magnitude. These lineaments are assumed to represent vertical faults. Smaller earthquakes are located at random. The magnitude of the earthquake that any fault could generate is based upon the mapped surface trace length of the lineaments, and is calculated from regression relations. Recurrence rates for a given magnitude of earthquake are based upon published studies for Sweden. A major assumption in this study is that future earthquakes will be similar in magnitude, location and orientation as earthquakes in the geological and historical records of Sweden. Another important assumption is that the displacement calculations based upon linear elasticity and linear elastic fracture mechanics provides a conservative (over-)estimate of possible displacements. A third assumption is that the world

  18. Syntheses, structure and magnetic properties of two vanadate garnets Ca{sub 5}M{sub 4}V{sub 6}O{sub 24} (M=Co, Ni)

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Nannan [State Key Laboratory of Structural Chemistry, Fujian Institute of Research on the Structure of Matter, Chinese Academy of Sciences, Fuzhou 350002 (China); College of Materials Science and Engineering, Fuzhou University, Fuzhou, Fujian 350108 (China); He, Zhangzhen, E-mail: hcz1988@hotmail.com [State Key Laboratory of Structural Chemistry, Fujian Institute of Research on the Structure of Matter, Chinese Academy of Sciences, Fuzhou 350002 (China); Key Laboratory of Optoelectronic Materials Chemistry and Physics, Fujian Institute of Research on the Structure of Matter, Chinese Academy of Sciences, Fuzhou 350002 (China); Cui, Meiyan; Guo, Wenbin; Zhang, Suyun; Yang, Ming; Tang, Yingying [State Key Laboratory of Structural Chemistry, Fujian Institute of Research on the Structure of Matter, Chinese Academy of Sciences, Fuzhou 350002 (China)

    2015-08-15

    Two vanadate compounds Ca{sub 5}M{sub 4}V{sub 6}O{sub 24} (M=Co, Ni) have been synthesized by a high-temperature solid-state reaction. The compounds are found to crystallize in the cubic system with a space group Ia-3d, which exhibit a typical garnet structural framework. Magnetic measurements show that Ca{sub 5}M{sub 4}V{sub 6}O{sub 24} (M=Co, Ni) exhibit similar magnetic behaviors, in which Ca{sub 5}Co{sub 4}V{sub 6}O{sub 24} possesses an antiferromagnetic ordering at T{sub N}=~6 K while Ca{sub 5}Ni{sub 4}V{sub 6}O{sub 24} shows an antiferromagnetic ordering at T{sub N}=~7 K. - Graphical abstract: Garnet vanadate compounds Ca{sub 5}M{sub 4}V{sub 6}O{sub 24} (M=Co, Ni) have been synthesized by a high-temperature solid-state reaction. Structural features and magnetic behaviors are also investigated. - Highlights: • New type of garnet vanadates Ca{sub 5}M{sub 4}V{sub 6}O{sub 24} (M=Co, Ni) are synthesized by a high-temperature solid-state reaction. • Structural features are confirmed by single crystal samples. • Magnetic behaviors are firstly investigated in the systems.

  19. NDE of copper canisters for long-term storage of spent nuclear fuel from the Swedish nuclear power plants

    Science.gov (United States)

    Stepinski, Tadeusz

    2003-07-01

    Sweden has been intensively developing methods for long term storage of spent fuel from the nuclear power plants for twenty-five years. A dedicated research program has been initiated and conducted by the Swedish company SKB (Swedish Nuclear Fuels and Waste Management Co.). After the interim storage SKB plans to encapsulate spent nuclear fuel in copper canisters that will be placed at a deep repository located in bedrock. The canisters filled with fuel rods will be sealed by an electron beam weld. This paper presents three complementary NDE techniques used for assessing the sealing weld in copper canisters, radiography, ultrasound, and eddy current. A powerful X-ray source and a digital detector are used for the radiography. An ultrasonic array system consisting of a phased ultrasonic array and a multi-channel electronics is used for the ultrasonic examination. The array system enables electronic focusing and rapid electronic scanning eliminating the use of a complicated mechanical scanner. A specially designed eddy current probe capable of detecting small voids at the depth up to 4 mm in copper is used for the eddy current inspection. Presently, all the NDE techniques are verified in SKB's Canister Laboratory where full scale canisters are welded and examined.

  20. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    D.R. Jackson; G.R. Kiebel

    1999-08-24

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

  1. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    2000-11-18

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed

  2. Filter Measurement System for Nuclear Material Storage Canisters. End of Year Report FY 2013

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-02-03

    A test system has been developed at Los Alamos National Laboratory to measure the aerosol collection efficiency of filters in the lids of storage canisters for special nuclear materials. Two FTS (filter test system) devices have been constructed; one will be used in the LANL TA-55 facility with lids from canisters that have stored nuclear material. The other FTS device will be used in TA-3 at the Radiation Protection Division’s Aerosol Engineering Facility. The TA-3 system will have an expanded analytical capability, compared to the TA-55 system that will be used for operational performance testing. The LANL FTS is intended to be automatic in operation, with independent instrument checks for each system component. The FTS has been described in a complete P&ID (piping and instrumentation diagram) sketch, included in this report. The TA-3 FTS system is currently in a proof-of-concept status, and TA-55 FTS is a production-quality prototype. The LANL specification for (Hagan and SAVY) storage canisters requires the filter shall “capture greater than 99.97% of 0.45-micron mean diameter dioctyl phthalate (DOP) aerosol at the rated flow with a DOP concentration of 65±15 micrograms per liter”. The percent penetration (PEN%) and pressure drop (DP) of fifteen (15) Hagan canister lids were measured by NFT Inc. (Golden, CO) over a period of time, starting in the year 2002. The Los Alamos FTS measured these quantities on June 21, 2013 and on Oct. 30, 2013. The LANL(6-21-2013) results did not statistically match the NFT Inc. data, and the LANL FTS system was re-evaluated, and the aerosol generator was replaced and the air flow measurement method was corrected. The subsequent LANL(10-30-2013) tests indicate that the PEN% results are statistically identical to the NFT Inc. results. The LANL(10-30-2013) pressure drop measurements are closer to the NFT Inc. data, but future work will be investigated. An operating procedure for the FTS (filter test system) was written, and

  3. Very deep borehole. Deutag's opinion on boring, canister emplacement and retrievability

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Tim [Well Engineering Partners BV, The Hague (Netherlands)

    2000-05-01

    An engineering feasibility study has been carried out to determine whether or not it is possible to drill the proposed Very Deep Borehole concept wells required by SKB for nuclear waste disposal. A conceptual well design has been proposed. All aspects of well design have been considered, including drilling tools, rig design, drilling fluids, casing design and annulus isolation. The proposed well design is for 1168.4 mm hole to be drilled to 500 m. A 1066.8 mm outer diameter (OD) casing will be run and cemented. A 1016 mm hole will be drilled to approximately 2000 m, where 914.4 mm OD casing will be run. This annulus will be sealed with bentonite slurry apart from the bottom 100 m which will be cemented. 838.2 mm hole will be drilled to a final depth of 4000 m, where 762 mm OD slotted casing will be run. All the hole sections will be drilled using a downhole hammer with foam as the drilling fluid medium. Prior to running each casing string, the hole will be displaced to mud to assist with casing running and cementing. The waste canisters will be run on a simple J-slot tool, with integral backup system in case the J-slot fails. The canisters will all be centralised. Canisters can be retrieved using the same tool as used to run them. Procedures are given for both running and retrieving. Logging and testing is recommended only in the exploratory wells, in a maximum hole size of 311.1 mm. This will require the drilling of pilot holes to enable logging and testing to take place. It is estimated that each well will take approximately 137 days to drill and case, at an estimated cost of 4.65 Meuro per well. This time and cost estimate does not include any logging, testing, pilot hole drilling or time taken to run the canisters. New technology developments to enhance the drilling process are required in recyclable foam systems, in hammer bit technology, and in the development of robust under-reamers. It is the authors conclusion that it is possible to drill the well with

  4. Effect of component substitution on the microstructure and mechanical properties of MCoCrFeNiTix (M = Cu,Al) solid-solution alloys

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    MCoCrFeNiTix (M = Cu,Al;x:molar ratio,x = 0,0.5) alloys were prepared using the new alloy-design strategy of equal-atomic ratio and high entropy.By the component substitution of Al for Cu,the microstructure changes from the face-centered cubic solid solution of original CuCoCrFeNiTix alloys to the body-centered cubic solid solution of AICoCrFeNiTix alloys.Compared with original CuCoCrFeNiTix alloys,AICoCrFeNiTix alloys keep the similar good ductility and simultaneously possess a much higher compressive strength,which are even superior to most of the reported high-strength alloys like bulk metallic glasses.

  5. HANSF 1.3 user's manual

    Energy Technology Data Exchange (ETDEWEB)

    PLYS, M.G.

    1999-05-21

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. It may be used for all phases of spent fuel disposition including cold vacuum drying, transportation, and storage. This manual reflects HANSF version 1.3, a revised version of version 1.2a. HANSF 1.3 was written to add new models for axial nodalization, add new features for ease of usage, and correct errors. HANSF 1.3 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under a DOS-type operating system. HANSF 1.3 is known to compile under Lahey TI and Digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments.

  6. Summary of canister overheating incident at the Carbon Tetrachloride Expedited Response Action site

    Energy Technology Data Exchange (ETDEWEB)

    Driggers, S.A.

    1994-03-10

    The granular activated carbon (GAC)-filled canister that overheated was being used to adsorb carbon tetrachloride vapors drawn from a well near the 216-Z-9 Trench, a subsurface disposal site in the 200 West Area of the Hanford Site. The overheating incident resulted in a band of discolored paint on the exterior surface of the canister. Although there was no other known damage to equipment, no injuries to operating personnel, and no releases of hazardous materials, the incident is of concern because it was not anticipated. It also poses the possibility of release of carbon tetrachloride and other hazardous vapors if the incident were to recur. All soil vapor extraction system (VES) operations were halted until a better understanding of the cause of the incident could be determined and controls implemented to reduce the possibility of a recurrence. The focus of this report and the intent of all the activities associated with understanding the overheating incident has been to provide information that will allow safe restart of the VES operations, develop operational limits and controls to prevent recurrence of an overheating incident, and safely optimize recovery of carbon tetrachloride from the ground.

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

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.

    1996-11-01

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

  8. Genesis Solar Wind Science Canister Components Curated as Potential Solar Wind Collectors and Reference Contamination Sources

    Science.gov (United States)

    Allton, J. H.; Gonzalez, C. P.; Allums, K. K.

    2016-01-01

    The Genesis mission collected solar wind for 27 months at Earth-Sun L1 on both passive and active collectors carried inside of a Science Canister, which was cleaned and assembled in an ISO Class 4 cleanroom prior to launch. The primary passive collectors, 271 individual hexagons and 30 half-hexagons of semiconductor materials, are described in. Since the hard landing reduced the 301 passive collectors to many thousand smaller fragments, characterization and posting in the online catalog remains a work in progress, with about 19% of the total area characterized to date. Other passive collectors, surfaces of opportunity, have been added to the online catalog. For species needing to be concentrated for precise measurement (e.g. oxygen and nitrogen isotopes) an energy-independent parabolic ion mirror focused ions onto a 6.2 cm diameter target. The target materials, as recovered after landing, are described in. The online catalog of these solar wind collectors, a work in progress, can be found at: http://curator.jsc.nasa.gov/gencatalog/index.cfm This paper describes the next step, the cataloging of pieces of the Science Canister, which were surfaces exposed to the solar wind or component materials adjacent to solar wind collectors which may have contributed contamination.

  9. Rates and mechanisms of radioactive release and retention inside a waste disposal canister - in Can Processes

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M. (ed.) [and others

    2003-10-01

    Sweden and Finland are planning to dispose of spent nuclear fuel in a deep underground repository constructed in granitic rock. Each country is investigating candidate sites and developing the scientific and technical basis for assessing the safety of an eventual repository. An essential part of the safety assessment involves understanding the behaviour of the spent fuel after it is placed in the geologic environment. The fuel will be sealed inside a copper canister that contains a cast iron insert. The copper functions as a corrosion resistant barrier, while the cast iron insert fills much of the internal void space, adding strength to the canister and reducing the space available for water to accumulate inside the canister after the corrosion barrier is breached. The canisters will be surrounded by compressed bentonite, which will limit the access of water and dissolved species to the canister. Oxygen that is initially present when the disposal environment is sealed will be rapidly consumed by pyrite in the bentonite, bacterial species in the rock, and reduced inorganic materials in the rock. The copper canister will prevent access of water to the iron until it is corroded through, a process that is expected to take millions of years. After water contacts the iron, anaerobic corrosion of the insert will generate hydrogen gas and introduce Fe(II) ions into the water. The long-term environment for the fuel, therefore, is a highly reducing environment. The only possible source of oxidising agents is radiolysis of the water by radiation from the fuel. In the long-term, the radioactivity in the fuel is due to isotopes that decay by alpha decay; most of the activity from beta and gamma radiation will have decayed away. Spent fuel that is available for testing contains high levels of beta and gamma activity. Even when testing is done in the presence of hydrogen or actively corroding iron, the radiolysis due to beta and gamma radiation can introduce oxidising agents into

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

    Energy Technology Data Exchange (ETDEWEB)

    Kaufhold, Stephan, E-mail: s.kaufhold@bgr.de [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); Hassel, Achim Walter [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Institute for Chemical Technology of Inorganic Materials, Johannes Kepler University Linz, Altenberger Straße 69, 4040 Linz (Austria); Sanders, Daniel [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Dohrmann, Reiner [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); LBEG, Landesamt für Bergbau, Energie und Geologie, Stilleweg 2, D-30655 Hannover (Germany)

    2015-03-21

    Graphical abstract: Corrosion at the bentonite iron interface proceeds unaerobically with formation of an 1:1 Fe silicate mineral. A series of exposure tests with different types of bentonites showed that Na–bentonites are slightly less corrosive than Ca–bentonites and highly charges smectites are less corrosive compared to low charged ones. The formation of a patina was observed in some cases and has to be investigated further. - Highlights: • At the iron bentonite interface a 1:1 Fe layer silicate forms upon corrosion. • A series of iron–bentonite corrosion products showed slightly less corrosion for Na-rich and high-charged bentonites. • In some tests the formation of a patina was observed consisting of Fe–silicate, which has to be investigated further. - Abstract: Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na–bentonites compared to the Ca–bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe

  11. Biogeochemistry of Redox at Repository Depth and Implications for the Canister

    Energy Technology Data Exchange (ETDEWEB)

    Bath, Adrian; Hermansson, Hans-Peter

    2009-08-15

    The present groundwater chemical conditions at the candidate sites for a spent nuclear fuel repository in Sweden (the Forsmark and Laxemar sites) and processes affecting its future evolution comprise essential conditions for the evaluation of barrier performance and long-term safety. This report reviews available chemical sampling information from the site investigations at the candidate sites, with a particular emphasis on redox active groundwater components and microbial populations that influence redox affecting components. Corrosion of copper canister material is the main barrier performance influence of redox conditions that is elaborated in the report. One section addresses native copper as a reasonable analogue for canister materials and another addresses the feasibility of methane hydrate ice accumulation during permafrost conditions. Such an accumulation could increase organic carbon availability in scenarios involving microbial sulphate reduction. The purpose of the project is to evaluate and describe the available knowledge and data for interpretation of geochemistry, microbiology and corrosion in safety assessment. A conclusive assessment of the sufficiency of information can, however, only be done in the future context of a full safety assessment. The authors conclude that SKB's data and models for chemical and microbial processes are adequate and reasonably coherent. The redox conditions in the repository horizon are predominantly established through the SO{sub 4}2-/HS- and Fe3+/Fe2+ redox couples. The former may exhibit a more significant buffering effect as suggested by measured Eh values, while the latter is associated with a lager capacity due to abundant Fe(II) minerals in the bedrock. Among a large numbers of groundwater features considered in geochemical equilibrium modelling, Eh, pH, temperature and concentration of dissolved sulphide comprise the most essential canister corrosion influences. Groundwater sulphide may originate from

  12. Galvanic and stress corrosion of copper canisters in repository environment. A short review

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P.; Koenig, M. [Studsvik Nuclear AB, Nykoeping (Sweden)

    2001-02-01

    The Swedish Nuclear Power Inspectorate, SKI, has studied different aspects of canister and copper corrosion as part of the general improvement of the knowledge base within the area. General and local corrosion has earlier been treated by experiments as well as by thermodynamic calculations. For completeness also galvanic and stress corrosion should be treated. The present work is a short review, intended to indicate areas needing further focus. The work consists of two parts, the first of which contains a judgement of statements concerning risk of galvanic corrosion of copper in the repository. The second part concerns threshold values for the stress intensity factor of stress corrosion in copper. A suggestion is given on how such values possibly could be measured for copper at repository conditions. In early investigations by SKB, galvanic corrosion is not mentioned or at least not treated. In later works it is treated but often in a theoretical way without indications of any further treatment or investigation. Several pieces of work indicate that further investigations are required to ensure that different types of corrosion, like galvanic, cannot occur in the repository environment. There are for example effects of grain size, grain boundary conditions, impurities and other factors that could influence the appearance of galvanic corrosion that are not treated. Those factors have to be considered to be completely sure that galvanic corrosion and related effects does not occur for the actual canister in the specific environment of the repository. The circumstances are so specific, that a rather general discussion indicating that galvanic corrosion is not probable just is not enough. Experiments should also be performed for verification. It is concluded that the following specific areas, amongst others, could benefit from further consideration. Galvanic corrosion of unbreached copper by inhomogeneities in the environment and in the copper metal should be addressed

  13. The Effect of Flow Rate and Canister Geometry on the Effectiveness of Removing Carbon Dioxide with Soda Lime.

    Science.gov (United States)

    1980-09-01

    pressure was measured using a Meriam type W,0 - 30 inches of mercury manometer . The gas was then piped to the water bath. At the water bath, the gas was...bypass pressure as indicatedon the 30-inch mercury manometer was recorded. The canister was then allowed to remain in the water bath for forty-five

  14. Canister storage building (CSB) safety analysis report phase 3: Safety analysis documentation supporting CSB construction

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1997-04-28

    The Canister Storage Building (CSB) will be constructed in the 200 East Area of the U.S. Department of Energy (DOE) Hanford Site. The CSB will be used to stage and store spent nuclear fuel (SNF) removed from the Hanford Site K Basins. The objective of this chapter is to describe the characteristics of the site on which the CSB will be located. This description will support the hazard analysis and accident analyses in Chapter 3.0. The purpose of this report is to provide an evaluation of the CSB design criteria, the design's compliance with the applicable criteria, and the basis for authorization to proceed with construction of the CSB.

  15. Site-to-canister scale flow and transport in Haestholmen, Kivetty, Olkiluoto and Romuvaara

    Energy Technology Data Exchange (ETDEWEB)

    Poteri, A.; Laitinen, M. [VTT Energy, Espoo (Finland)

    1999-05-01

    Radioactive waste is originating from production of electricity in nuclear power plants. Most of the waste has only low or intermediate levels of radioactivity. However, the spent nuclear fuel is highly radioactive and it has to be isolated from the biosphere. The current nuclear waste management plan in Finland is based on direct disposal of the spent nuclear fuel deep underground. The only feasible mechanism for the radionuclides to escape from an underground repository is to be carried by the groundwater flow after the failure of waste containers. The scope of this study is to examine the groundwater flow situation and transport properties in the vicinity of the disposal canister and along the potential release paths from the repository into the biosphere. The results of this study are further applied in the site specific safety analysis of a spent fuel repository. Synthesis is made of the porous medium estimates of the groundwater flow in the regional and site scales and the detailed fracture network analysis of the flow in the canister scale. This synthesis includes estimation of the transport properties from the canister into the biosphere and flow rates around the deposition holes of the waste canisters. The modelling has been carried out for four different sites: Hastholmen, Kivetty, Olkiluoto and Romavaara. According to the simulations groundwater flow rate around the deposition holes is less than about 1 litre/a for about 75 % of the deposition holes. For about 5 % of the deposition holes the flow rates are a few litres per year or higher. The highest flow rates resulted at Hastholmen, in fresh water conditions 10 000 years after present, and at Kivetty. The transport resistances were calculated for the `worst` flow paths that might have impact on the safety of the repository. The total transport resistances from the repository into the biosphere along those flow paths varied between about 40 000 a/m and 5-10{sup 6} a/m. Most of the total transport

  16. Fuel and canister process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars (ed.)

    2006-10-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel.

  17. Evaluation of Ca3(Co,M2O6 (M=Co, Fe, Mn, Ni as new cathode materials for solid-oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Fushao Li

    2015-10-01

    Full Text Available Series compounds Ca3(Co0.9M0.12O6 (M=Co, Fe, Mn, Ni with hexagonal crystal structure were prepared by sol–gel route as the cathode materials for solid oxide fuel cells (SOFCs. Effects of the varied atomic compositions on the structure, electrical conductivity, thermal expansion and electrochemical performance were systematically evaluated. Experimental results showed that the lattice parameters of Ca3(Co0.9Fe0.12O6 and Ca3(Co0.9Mn0.12O6 were both expanded to certain degree. Electron-doping and hole-doping effects were expected in Ca3(Co0.9Mn0.12O6 and Ca3(Co0.9Ni0.12O6 respectively according to the chemical states of constituent elements and thermal-activated behavior of electrical conductivity. Thermal expansion coefficients (TEC of Ca3(Co0.9M0.12O6 were measured to be distributed around 16×10−6 K−1, and compositional elements of Fe, Mn, and Ni were especially beneficial for alleviation of the thermal expansion problem of cathode materials. By using Ca3(Co0.9M0.12O6 as the cathodes operated at 800 °C, the interfacial area-specific resistance varied in the order of M=CoM=Co, Fe, Mn, Ni can be used as the cost-effective cathode materials for SOFCs.

  18. Creep of the Copper Canister. A Critical Review of the Literature

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, William H. [Meadow End Farm, Farnham (United Kingdom)

    2003-04-01

    Literature relevant to creep of the copper shell of the copper-iron canister has been reviewed. Two classes of copper have been examined, Oxygen Free High Conductivity (OFHC), which is referred to in the relevant literature and this report as OF material, and OF material with 50 ppm of phosphorus added. The second material is referred to as OFP. Creep processes occurring in copper are briefly described and a deformation diagram, after Frost and Ashby is provided. It is concluded that the diagram adequately describes the processes observed for the two materials of interest without necessarily being in exact agreement at a quantitative level. There are two regimes of time, temperature and stress which are important when creep of the copper shell is considered. The first is a holding period between welding of the lid to the canister and placing the canister in the repository and the second is the storage period in the repository. In the holding period, residual stresses arising from the manufacturing processes are important and in the second period stresses arising from repository pressures are important as well as the residual pressures arising from manufacture. The holding period may extend up to one year and the temperature of the copper shell may decline from the immediate post welding temperature to 100 deg C in this interval. Initial peak localised stresses may give rise to strains of up to 14 %. Dynamic recovery immediately after welding reduces the stresses associated with these strains to levels which correspond to stresses for approximately 0.1 % strain at the ruling temperature. This is 75 MPa at 100 deg C and 50 MPa for 150 deg C. A further stress relaxation of up to 30 % occurs in the first 20 days after welding. Localised stresses are therefore unlikely to exceed 50 MPa when the canister is placed into storage. No negative effects have been observed in connection with this stress relaxation process. In the storage period, which is indefinite, the

  19. Curcumin as the OO bidentate ligand in "2 + 1" complexes with the [M(CO)3]+ (M = Re, 99mTc) tricarbonyl core for radiodiagnostic applications.

    Science.gov (United States)

    Sagnou, Marina; Benaki, Dimitra; Triantis, Charalampos; Tsotakos, Theodoros; Psycharis, Vassilis; Raptopoulou, Catherine P; Pirmettis, Ioannis; Papadopoulos, Minas; Pelecanou, Maria

    2011-02-21

    The synthesis and characterization of "2 + 1" complexes of the [M(CO)(3)](+) (M = Re, (99m)Tc) core with the β-diketones acetylacetone (complexes 2, 8) and curcumin (complexes 5, 10 and 6, 11) as bidentate OO ligands, and imidazole or isocyanocyclohexane as monodentate ligands is reported. The complexes were synthesized by reacting the [NEt(4)](2)[Re(CO)(3)Br(3)] precursor with the β-diketone to generate the intermediate aqua complex fac-Re(CO)(3)(OO)(H(2)O) that was isolated and characterized, followed by replacement of the labile water by the monodentate ligand. All complexes were characterized by mass spectrometry, NMR and IR spectroscopies, and elemental analysis. In the case of complex 2, bearing imidazole as the monodentate ligand, X-ray analysis was possible. The chemistry was successfully transferred at (99m)Tc tracer level. The curcumin complexes 5 and 6, as well as their intermediate aqua complex 4, that bear potential for radiopharmaceutical applications due to the wide spectrum of pharmacological activity of curcumin, were successfully tested for selective staining of β-amyloid plaques of Alzheimer's disease. The fact that the complexes maintain the affinity of the mother compound curcumin for β-amyloid plaques prompts for further exploration of their chemistry and biological properties as radioimaging probes.

  20. A DFT Study on Selected Physical Organic Aspects of the Fischer Carbene Intermediates [(M(CO4(C(OMeMe

    Directory of Open Access Journals (Sweden)

    Tareq Irshaidat

    2010-01-01

    Full Text Available Fischer carbenes are important starting materials for C-C bond formation via coupling reactions between carbene and wide variety of substituted alkenes or alkynes. This DFT study shed light on unique fundamental organic/organometallic aspects for the C(OMeMe carbene in the free form and in case of bonding with M(CO4 (M= Cr, Mo, W. The data illustrate that the structures of the title intermediates include a unique structure stabilizing intramolecular M…C-H interaction (agostic interaction. This conclusion was made based on calculated NMR data (for carbon and hydrogen, structural parameters, energy calculations of conformers (C-C conformation, selected IR stretching frequencies (C-O, C-C, and C-H, and atomic charges. The agostic interaction is most efficient in case of chromium and in general is described as an overlap between the σ-bond electron pair of C-H with an empty d-orbital of the metal. These characterized examples are new addition to the orbital interaction theory.

  1. Fermi surfaces and Phase Stability of Ba(Fe$_{1-x}$M$_x$)$_2$As$_2$ (M=Co, Ni, Cu, Zn)

    CERN Document Server

    Khan, Suffian; Johnson, Duane

    2014-01-01

    BaFe$_2$As$_2$ with transition-metal doping exhibits a variety of rich phenomenon from coupling of structure, magnetism, and superconductivity. Using density functional theory, we systematically compare the Fermi surfaces (FS), formation energies ($\\Delta E_f$), and density of states (DOS) of electron-doped Ba(Fe$_{1-x}$M$_x$)$_2$As$_2$ with M={Co, Ni, Cu, Zn} in tetragonal (I$4/mmm$) and orthorhombic (F$mmm$) structures in nonmagnetic (NM), antiferromagnetic (AFM), and paramagnetic (PM, disordered local moment) states. We explain changes to phase stability ($\\Delta E_f$) and Fermi surfaces (and nesting) due to chemical and magnetic disorder, and compare to observed/assessed properties and contrast alloy theory with that expected from rigid-band model. With alloying, the DOS changes from common-band (Co,Ni) to split-band (Cu,Zn), which dictates $\\Delta E_f$ and can overwhelm FS-nesting instabilities, as for Cu,Zn cases.

  2. Final Report: Part 1. In-Place Filter Testing Instrument for Nuclear Material Containers. Part 2. Canister Filter Test Standards for Aerosol Capture Rates.

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Austin Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Runnels, Joel T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-11-02

    A portable instrument has been developed to assess the functionality of filter sand o-rings on nuclear material storage canisters, without requiring removal of the canister lid. Additionally, a set of fifteen filter standards were procured for verifying aerosol leakage and pressure drop measurements in the Los Alamos Filter Test System. The US Department of Energy uses several thousand canisters for storing nuclear material in different chemical and physical forms. Specialized filters are installed into canister lids to allow gases to escape, and to maintain an internal ambient pressure while containing radioactive contaminants. Diagnosing the condition of container filters and canister integrity is important to ensure worker and public safety and for determining the handling requirements of legacy apparatus. This report describes the In-Place-Filter-Tester, the Instrument Development Plan and the Instrument Operating Method that were developed at the Los Alamos National Laboratory to determine the “as found” condition of unopened storage canisters. The Instrument Operating Method provides instructions for future evaluations of as-found canisters packaged with nuclear material. Customized stainless steel canister interfaces were developed for pressure-port access and to apply a suction clamping force for the interface. These are compatible with selected Hagan-style and SAVY-4000 storage canisters that were purchased from NFT (Nuclear Filter Technology, Golden, CO). Two instruments were developed for this effort: an initial Los Alamos POC (Proof-of-Concept) unit and the final Los Alamos IPFT system. The Los Alamos POC was used to create the Instrument Development Plan: (1) to determine the air flow and pressure characteristics associated with canister filter clogging, and (2) to test simulated configurations that mimicked canister leakage paths. The canister leakage scenarios included quantifying: (A) air leakage due to foreign material (i.e. dust and hair

  3. Evaluation of the conservativeness of the methodology for estimating earthquake-induced movements of fractures intersecting canisters

    Energy Technology Data Exchange (ETDEWEB)

    La Pointe, Paul R.; Cladouhos, Trenton T. [Golder Associates Inc., Las Vegas, NV (United States); Outters, Nils; Follin, Sven [Golder Grundteknik KB, Stockholm (Sweden)

    2000-04-01

    This study evaluates the parameter sensitivity and the conservativeness of the methodology outlined in TR 99-03. Sensitivity analysis focuses on understanding how variability in input parameter values impacts the calculated fracture displacements. These studies clarify what parameters play the greatest role in fracture movements, and help define critical values of these parameters in terms of canister failures. The thresholds or intervals of values that lead to a certain level of canister failure calculated in this study could be useful for evaluating future candidate sites. Key parameters include: 1. magnitude/frequency of earthquakes; 2. the distance of the earthquake from the canisters; 3. the size and aspect ratio of fractures intersecting canisters; and 4. the orientation of the fractures. The results of this study show that distance and earthquake magnitude are the most important factors, followed by fracture size. Fracture orientation is much less important. Regression relations were developed to predict induced fracture slip as a function of distance and either earthquake magnitude or slip on the earthquake fault. These regression relations were validated by using them to estimate the number of canister failures due to single damaging earthquakes at Aberg, and comparing these estimates with those presented in TR 99-03. The methodology described in TR 99-03 employs several conservative simplifications in order to devise a numerically feasible method to estimate fracture movements due to earthquakes outside of the repository over the next 100,000 years. These simplifications include: 1. fractures are assumed to be frictionless and cohesionless; 2. all energy transmitted to the fracture by the earthquake is assumed to produce elastic deformation of the fracture; no energy is diverted into fracture propagation; and 3. shielding effects of other fractures between the earthquake and the fracture are neglected. The numerical modeling effectively assumes that the

  4. Modeling of molecular and particulate transport in dry spent nuclear fuel canisters

    Science.gov (United States)

    Casella, Andrew M.

    2007-09-01

    The transportation and storage of spent nuclear fuel is one of the prominent issues facing the commercial nuclear industry today, as there is still no general consensus regarding the near- and long-term strategy for managing the back-end of the nuclear fuel cycle. The debate continues over whether the fuel cycle should remain open, in which case spent fuel will be stored at on-site reactor facilities, interim facilities, or a geologic repository; or if the fuel cycle should be closed, in which case spent fuel will be recycled. Currently, commercial spent nuclear fuel is stored at on-site reactor facilities either in pools or in dry storage containers. Increasingly, spent fuel is being moved to dry storage containers due to decreased costs relative to pools. As the number of dry spent fuel containers increases and the roles they play in the nuclear fuel cycle increase, more regulations will be enacted to ensure that they function properly. Accordingly, they will have to be carefully analyzed for normal conditions, as well as any off-normal conditions of concern. This thesis addresses the phenomena associated with one such concern; the formation of a microscopic through-wall breach in a dry storage container. Particular emphasis is placed on the depressurization of the canister, release of radioactivity, and plugging of the breach due to deposition of suspended particulates. The depressurization of a dry storage container upon the formation of a breach depends on the temperature and quantity of the fill gas, the pressure differential across the breach, and the size of the breach. The first model constructed in this thesis is capable of determining the depressurization time for a breached container as long as the associated parameters just identified allow for laminar flow through the breach. The parameters can be manipulated to quantitatively determine their effect on depressurization. This model is expanded to account for the presence of suspended particles. If

  5. End of FY2014 Report - Filter Measurement System for Nuclear Material Storage Canisters (Including Altitude Correction for Filter Pressure Drop)

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-24

    Two LANL FTS (Filter Test System ) devices for nuclear material storage canisters are fully operational. One is located in PF-4 ( i.e. the TA-55 FTS) while the other is located at the Radiation Protection Division’s Aerosol Engineering Facility ( i.e. the TA-3 FTS). The systems are functionally equivalent , with the TA-3 FTS being the test-bed for new additions and for resolving any issues found in the TA-55 FTS. There is currently one unresolved issue regarding the TA-55 FTS device. The canister lid clamp does not give a leak tight seal when testing the 1 QT (quart) or 2 QT SAVY lids. An adapter plate is being developed that will ensure a correct test configuration when the 1 or 2 QT SAVY lid s are being tested .

  6. Report on hydro-mechanical and chemical-mineralogical analyses of the bentonite buffer in Canister Retrieval Test

    Energy Technology Data Exchange (ETDEWEB)

    Dueck, Ann; Johannesson, Lars-Erik; Kristensson, Ola; Olsson, Siv [Clay Technology AB (Sweden)

    2011-12-15

    The effect of five years of exposure to repository-like conditions on compacted Wyoming bentonite was determined by comparing the hydraulic, mechanical, and mineralogical properties of samples from the bentonite buffer of the Canister Retrieval Test (CRT) with those of reference material. The CRT, located at the Swedish Aspo Hard Rock Laboratory (HRL), was a full-scale field experiment simulating conditions relevant for the Swedish KBS-3 concept for disposal of high-level radioactive waste in crystalline host rock. The compacted bentonite, surrounding a copper canister equipped with heaters, had been subjected to heating at temperatures up to 95 deg C and hydration by natural Na-Ca-Cl type groundwater for almost five years at the time of retrieval. Under the thermal and hydration gradients that prevailed during the test, sulfate in the bentonite was redistributed and accumulated as anhydrite close to the canister. The major change in the exchangeable cation pool was a loss in Mg in the outer parts of the blocks, suggesting replacement of Mg mainly by Ca along with the hydration with groundwater. Close to the copper canister, small amounts of Cu were incorporated in the bentonite. A reduction of strain at failure was observed in the innermost part of the bentonite buffer, but no influence was seen on the shear strength. No change of the swelling pressure was observed, while a modest decrease in hydraulic conductivity was found for the samples with the highest densities. No coupling was found between these changes in the hydro-mechanical properties and the montmorillonite . the X-ray diffraction characteristics, the cation exchange properties, and the average crystal chemistry of the Na-converted < 1 {mu}m fractions provided no evidence of any chemical/structural changes in the montmorillonite after the 5-year hydrothermal test.

  7. INITIAL WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: MULTI-PURPOSE CANISTER WITH DISPOSAL CONTAINER (TBV)

    Energy Technology Data Exchange (ETDEWEB)

    J.R. Massari

    1995-10-06

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide an assessment of the present waste package design from a criticality risk standpoint. The specific objectives of this initial analysis are to: (1) Establish a process for determining the probability of waste package criticality as a function of time (in terms of a cumulative distribution function, probability distribution function, or expected number of criticalities in a specified time interval) for various waste package concepts; (2) Demonstrate the established process by estimating the probability of criticality as a function of time since emplacement for an intact multi-purpose canister waste package (MPC-WP) configuration; (3) Identify the dominant sequences leading to waste package criticality for subsequent detailed analysis. The purpose of this analysis is to document and demonstrate the developed process as it has been applied to the MPC-WP. This revision is performed to correct deficiencies in the previous revision and provide further detail on the calculations performed. This analysis is similar to that performed for the uncanistered fuel waste package (UCF-WP, B00000000-01717-2200-00079).

  8. Human Factors Engineering and Ergonomics Analysis for the Canister Storage Building (CSB) Results and Findings

    Energy Technology Data Exchange (ETDEWEB)

    GARVIN, L.J.

    1999-09-20

    The purpose for this supplemental report is to follow-up and update the information in SNF-3907, Human Factors Engineering (HFE) Analysis: Results and Findings. This supplemental report responds to applicable U.S. Department of Energy Safety Analysis Report review team comments and questions. This Human Factors Engineering and Ergonomics (HFE/Erg) analysis was conducted from April 1999 to July 1999; SNF-3907 was based on analyses accomplished in October 1998. The HFE/Erg findings presented in this report and SNF-3907, along with the results of HNF-3553, Spent Nuclear Fuel Project, Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report,'' Chapter A3.0, ''Hazards and Accidents Analyses,'' provide the technical basis for preparing or updating HNF-3553. Annex A, Chaptex A13.0, ''Human Factors Engineering.'' The findings presented in this report allow the HNF-3553 Chapter 13.0, ''Human Factors,'' to respond fully to the HFE requirements established in DOE Order 5480.23, Nuclear Safety Analysis Reports.

  9. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Algorithms for ultrasonic imaging

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Engholm, Marcus; Olofsson, Tomas (Uppsala Univ., Signals and Systems, Dept. of Technical Sciences (Sweden))

    2011-07-15

    This report contains research results concerning the use of advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala Univ. in 2009 and 2010. The first part of the report deals with ultrasonic imaging of damage in planar structures using Lamb waves. We present results of the first successful attempt to apply an adaptive beamformer for Lamb waves. Our algorithm is an extension of the adaptive beamformer based on minimum variance distortion less response (MVDR) approach to dispersive, multimodal Lamb waves. We present simulation and experimental results illustrating the performance of the MVDR applied to imaging artificial damage in an aluminum plate. In the second part of the report we present two extensions of the previously proposed 2D phase shift migration algorithms for enhancing resolution in ultrasonic imaging of solid objects. The first extension enables processing 3D data in order to fully utilize the resolution enhancement potential of the technique. The second extension, consists in generalizing the technique to allow for the processing of data acquired using an array instead of a previously concerned single transducer. Robustness issue related to objects having front surfaces that are slightly tilted relative to the scanning axis is also considered

  10. A methodology to estimate earthquake effects on fractures intersecting canister holes

    Energy Technology Data Exchange (ETDEWEB)

    La Pointe, P.; Wallmann, P.; Thomas, A.; Follin, S. [Golder Assocites Inc. (Sweden)

    1997-03-01

    A literature review and a preliminary numerical modeling study were carried out to develop and demonstrate a method for estimating displacements on fractures near to or intersecting canister emplacement holes. The method can be applied during preliminary evaluation of candidate sites prior to any detailed drilling or underground excavation, utilizing lineament maps and published regression relations between surface rupture trace length and earthquake magnitude, rupture area and displacements. The calculated displacements can be applied to lineament traces which are assumed to be faults and may be the sites for future earthquakes. Next, a discrete fracture model is created for secondary faulting and jointing in the vicinity of the repository. These secondary fractures may displace due to the earthquake on the primary faults. The three-dimensional numerical model assumes linear elasticity and linear elastic fracture mechanics which provides a conservative displacement estimate, while still preserving realistic fracture patterns. Two series of numerical studies were undertaken to demonstrate how the methodology could be implemented and how results could be applied to questions regarding site selection and performance assessment. The first series illustrates how earthquake damage to a hypothetical repository for a specified location (Aespoe) could be estimated. A second series examined the displacements induced by earthquakes varying in magnitude from 6.0 to 8.2 as a function of how close the earthquake was in relation to the repository. 143 refs, 25 figs, 7 tabs.

  11. 金属氢化物储氢装置研究%Study on Metal Hydride Canister

    Institute of Scientific and Technical Information of China (English)

    刘晓鹏; 蒋利军; 陈立新

    2009-01-01

    The temperature field of the inner cylindrical canister was simulated by using finite difference method and 2D heat transfer model during the hydrogenation process. It is showed that a temperature gradient is distributed obviously in the metal hydride bed, and the ceutric place of the canister has the highest temperature. Therefore, heat assembled in the centric place must be intensively transferred to improve the hydrogen storage properties of the metal hydride canister. In order to improve the hydrogen absorption/desorption cycle performance of the canister, the cycle life of as-cast and melt-spinning TiV0.41 Fe0.09Mn1.5 alloy was comparatively studied. It is indicated that the cycle life of the melt -spinning alloy is considerably longer than that of the as-cast one. The canister prepared by using melt-spinning TiV0.41Fe0.09Mn1.5 alloy has 94% of the hydrogen storage capacity after 3600 cycles.%用有限差分法和二维导热模型计算了圆柱形金属氢化物储氢装置内部储氢过程的温度场分布,结果表明空气换热型储氢装置内部的合金反应床存在明显的温度梯度场,吸氢时储氢装置中心部位的温度最高,需要强化其芯部换热条件,以提高储氢装置的储放氢性能.对比研究了铸态以及甩带快淬工艺制备 TiV0.41 Fe0.09Mn1.5合金吸放氢循环寿命,表明甩带快淬工艺可以显著提高储氢合金的吸放氢循环性能.以甩带快淬工艺制备的TiV0.41Fe0.09Mn1.5合金为工质的储氢装置,经过3 600次吸放氢循环后的容量保持率达到94%以上.

  12. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, H. [Outokumpu Oy Poricopper, Pori (Finland)

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program `Weld 2000` and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  13. Syntheses, structure and magnetic properties of pillared layered diphosphonates: M2(O 3PC 6H 4PO 3)(H 2O) 2 ( M=Co II, Ni II)

    Science.gov (United States)

    Cao, Deng-Ke; Gao, Song; Zheng, Li-Min

    2004-07-01

    This paper describes the hydrothermal syntheses of two isostructural metal bisphosphonates: M2(O 3PC 6H 4PO 3)(H 2O) 2 [ M=Co II ( 1), Ni II ( 2)]. Single-crystal structure determination of compound 1 revealed a pillared layered structure in which the phenyl groups connect the inorganic layers of cobalt phosphonate. Crystal data for 1: orthorhombic, space group Pnnm, a=19.306(5), b=4.8293(12), c=5.6390(14) Å, V=525.7(2) Å 3, Z=2. Magnetic susceptibility data indicate that antiferromagnetic interactions are mediated in both cases.

  14. Analysis of the effect of vibrations on the bentonite buffer in the canister hole

    Energy Technology Data Exchange (ETDEWEB)

    Jonsson, Martin (AaF- Berg och Maetteknik, Stockholm (Sweden)); Hakami, Hossein; Ekneligoda, Thushan (Itasca Geomekanik AB, Solna (Sweden))

    2009-09-15

    During the construction of a final repository for spent nuclear fuel in crystalline rock, blasting activities in certain deposition tunnels will occur at the same time as the deposition of canisters containing the waste is going on in another adjacent access tunnel. In fact, the deposition consists of several stages after the drilling of the deposition hole. The most vulnerable stage from a vibration point of view is when the bentonite buffer is placed in the deposition hole but the canister has not been placed yet. During this stage, a hollow column of bentonite blocks remains free to vibrate inside the deposition hole. The goal of this study was to investigate the displacement of the bentonite blocks when exposed to the highest vibration level that can be expected during the drill and blast operations. In order to investigate this, a three dimensional model in 3DEC, capable of capturing the dynamic behaviour of the bentonite buffer was set up. To define the vibration levels, which serve as input data for the 3DEC model, an extensive analysis of the recorded vibrations from the TASQ - tunnel was carried out. For this purpose, an upper expected vibration limit was defined. This was done outgoing from the fact that the planned charging for the construction of the geological repository will lie in the interval 2 to 4 kg. Furthermore, at the first stage for this study, it was decided that the vibration should be conservatively evaluated for 30 m distance. Using these data, it was concluded that the maximum vibration level that can be expected will be approximately 60 mm/s. After simplifying the vibration signal, a sinusoidal wave with the amplitude 60 mm/s was applied at the bottom of the column and it was assumed that the vibrations only affect the bentonite buffer in one direction (horizontal direction). From this simulation, it was concluded that hardly any displacements occurred. However, when applying the same sinusoidal wave both in the horizontal and the

  15. Strategy for verification and demonstration of the sealing process for canisters for spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Christina [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany); Oeberg, Tomas [Tomas Oeberg Konsult AB, Lyckeby (Sweden)

    2004-08-01

    Electron beam welding and friction stir welding are the two processes now being considered for sealing copper canisters with Sweden's radioactive waste. This report outlines a strategy for verification and demonstration of the encapsulation process which here is considered to consist of the sealing of the canister by welding followed by quality control of the weld by non-destructive testing. Statistical methodology provides a firm basis for modern quality technology and design of experiments has been successful part of it. Factorial and fractional factorial designs can be used to evaluate main process factors and their interactions. Response surface methodology with multilevel designs enables further optimisation. Empirical polynomial models can through Taylor series expansions approximate the true underlying relationships sufficiently well. The fitting of response measurements is based on ordinary least squares regression or generalised linear methods. Unusual events, like failures in the lid welds, are best described with extreme value statistics and the extreme value paradigm give a rationale for extrapolation. Models based on block maxima (the generalised extreme value distribution) and peaks over threshold (the generalised Pareto distribution) are considered. Experiences from other fields of the materials sciences suggest that both of these approaches are useful. The initial verification experiments of the two welding technologies considered are suggested to proceed by experimental plans that can be accomplished with only four complete lid welds each. Similar experimental arrangements can be used to evaluate process 'robustness' and optimisation of the process window. Two series of twenty demonstration trials each, mimicking assembly-line production, are suggested as a final evaluation before the selection of welding technology. This demonstration is also expected to provide a data base suitable for a baseline estimate of future performance

  16. Cleaning Genesis Sample Return Canister for Flight: Lessons for Planetary Sample Return

    Science.gov (United States)

    Allton, J. H.; Hittle, J. D.; Mickelson, E. T.; Stansbery, Eileen K.

    2016-01-01

    Sample return missions require chemical contamination to be minimized and potential sources of contamination to be documented and preserved for future use. Genesis focused on and successfully accomplished the following: - Early involvement provided input to mission design: a) cleanable materials and cleanable design; b) mission operation parameters to minimize contamination during flight. - Established contamination control authority at a high level and developed knowledge and respect for contamination control across all institutions at the working level. - Provided state-of-the-art spacecraft assembly cleanroom facilities for science canister assembly and function testing. Both particulate and airborne molecular contamination was minimized. - Using ultrapure water, cleaned spacecraft components to a very high level. Stainless steel components were cleaned to carbon monolayer levels (10 (sup 15) carbon atoms per square centimeter). - Established long-term curation facility Lessons learned and areas for improvement, include: - Bare aluminum is not a cleanable surface and should not be used for components requiring extreme levels of cleanliness. The problem is formation of oxides during rigorous cleaning. - Representative coupons of relevant spacecraft components (cut from the same block at the same time with identical surface finish and cleaning history) should be acquired, documented and preserved. Genesis experience suggests that creation of these coupons would be facilitated by specification on the engineering component drawings. - Component handling history is critical for interpretation of analytical results on returned samples. This set of relevant documents is not the same as typical documentation for one-way missions and does include data from several institutions, which need to be unified. Dedicated resources need to be provided for acquiring and archiving appropriate documents in one location with easy access for decades. - Dedicated, knowledgeable

  17. Spin-Coating and Characterization of Multiferroic MFe{sub 2}O{sub 4} (M=Co, Ni) / BaTiO{sub 3} Bilayers

    Energy Technology Data Exchange (ETDEWEB)

    Quandt, Norman [Institute of Chemistry, Martin Luther University Halle-Wittenberg, Kurt-Mothes-Straße 2, 06120 Halle (Germany); Roth, Robert [Institute of Physics, Martin Luther University Halle-Wittenberg, Von-Danckelmann-Platz 3, 06120 Halle (Germany); Syrowatka, Frank [Interdisciplinary Center of Materials Science, Martin Luther University Halle-Wittenberg, Heinrich-Damerow-Straße 4, 06120 Halle (Germany); Steimecke, Matthias [Institute of Chemistry, Martin Luther University Halle-Wittenberg, Von-Danckelmann-Platz 4, 06120 Halle (Germany); Ebbinghaus, Stefan G., E-mail: stefan.ebbinghaus@chemie.uni-halle.de [Institute of Chemistry, Martin Luther University Halle-Wittenberg, Kurt-Mothes-Straße 2, 06120 Halle (Germany)

    2016-01-15

    Bilayer films of MFe{sub 2}O{sub 4} (M=Co, Ni) and BaTiO{sub 3} were prepared by spin coating of N,N-dimethylformamide/acetic acid solutions on platinum coated silicon wafers. Five coating steps were applied to get the desired thickness of 150 nm for both the ferrite and perovskite layer. XRD, IR and Raman spectroscopy revealed the formation of phase-pure ferrite spinels and BaTiO{sub 3}. Smooth surfaces with roughnesses in the order of 3 to 5 nm were found in AFM investigations. Saturation magnetization of 347 emu cm{sup −3} for the CoFe{sub 2}O{sub 4}/BaTiO{sub 3} and 188 emu cm{sup −3} for the NiFe{sub 2}O{sub 4}/BaTiO{sub 3} bilayer, respectively were found. For the CoFe{sub 2}O{sub 4}/BaTiO{sub 3} bilayer a strong magnetic anisotropy was observed with coercivity fields of 5.1 kOe and 3.3 kOe (applied magnetic field perpendicular and parallel to film surface), while for the NiFe{sub 2}O{sub 4}/BaTiO{sub 3} bilayer this effect is less pronounced. Saturated polarization hysteresis loops prove the presence of ferroelectricity in both systems. - Graphical abstract: The SEM image of the CoFe{sub 2}O{sub 4}/BaTiO{sub 3} bilayer on Pt–Si-substrate (left), magnetization as a function of the magnetic field perpendicular and parallel to the film plane (right top) and P–E and I–V hysteresis loops of the bilayer at room temperature. - Highlights: • Ferrite and perovskite oxides grown on platinum using spin coating technique. • Columnar growth of cobalt ferrite particle on the substrate. • Surface investigation showed a homogenous and smooth surface. • Perpendicular and parallel applied magnetic field revealed a magnetic anisotropy. • Switching peaks and saturated P–E hysteresis loops show ferroelectricity.

  18. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Ultrasonic imaging, FSW monitoring with acoustic emission

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Olofsson, Tomas; Wennerstroem, Erik [Uppsala Univ., Dept. of Technical Sciences (Sweden). Signals and Systems

    2006-12-15

    This report contains the research results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in years 2005/2006. In the first part of the report we propose a concept of monitoring of the friction stir welding (FSW) process by means of acoustic emission (AE) technique. First, we introduce the AE technique and then we present the principle of the system for monitoring the FSW process in cylindrical symmetry specific for the SKB canisters. We propose an omnidirectional circular array of ultrasonic transducers for receiving the AE signals generated by the FSW tool and the releases of the residual stress at canister's circumference. Finally, we review the theory of uniform circular arrays. The second part of the report is concerned with synthetic aperture focusing technique (SAFT) characterized by enhanced spatial resolution. We evaluate three different approaches to perform imaging with less computational cost than that of the extended SAFT (ESAFT) method proposed in our previous reports. First, a sparse version of ESAFT is presented, which solves the reconstruction problem only for a small set of the most probable scatterers in the image. A frequency domain the {omega}-k SAFT algorithm, which relies on the far-field approximation is presented in the second part. Finally, a detailed analysis of the most computationally intense step in the ESAFT and the sparse 2D deconvolution is presented. In the final part of the report we introduce basics of the 3D ultrasonic imaging that has a great potential in the inspection of the FSW welds. We discuss in some detail the three interrelated steps involved in the 3D ultrasonic imaging: data acquisition, 3D reconstruction, and 3D visualization.

  19. Fire simulation of the canister transfer and installation vehicle; Kapselin siirto- ja asennusajoneuvon palosimulointi

    Energy Technology Data Exchange (ETDEWEB)

    Peltokorpi, L. [Fortum Power and Heat Oy, Espoo (Finland)

    2012-12-15

    A pyrolysis model of the canister transfer and installation vehicle was developed and vehicle fires in the final disposal tunnel and in the central tunnel were simulated using the fire simulation program FDS (Fire Dynamics Simulator). For comparison, same vehicle fire was also simulated at conditions in which the fire remained as a fuel controlled during the whole simulation. The purpose of the fire simulations was to simulate the fire behaviour realistically taking into account for example the limitations coming from the lack of oxygen. The material parameters for the rubber were defined and the simulation models for the tyres developed by simulating the fire test of a front wheel loader rubber tyre done by SP Technical Research Institute of Sweden. In these simulations the most important phenomena were successfully brought out but the timing of the phenomena was difficult. The final values for the rubber material parameters were chosen so that the simulated fire behaviour was at least as intense as the measured one. In the vehicle fire simulations a hydraulic oil or diesel leak causing a pool fire size of 2 MW and 2 m{sup 2} was assumed. The pool fire was assumed to be located under the tyres of the SPMT (Self Propelled Modular Transporters) transporter. In each of the vehicle fire simulations only the tyres of the SPMT transporter were observed to be burning whereas the tyres of the trailer remained untouched. In the fuel controlled fire the maximum power was slightly under 10 MW which was reached in about 18 minutes. In the final disposal tunnel the growth of the fire was limited due to the lack of oxygen and the relatively fast air flows existing in the tunnel. Fast air flows caused the flame spreading to be limited to the certain directions. In the final disposal tunnel fire the maximum power was slightly over 7 MW which was reached about 8 minutes after the ignition. In the central tunnel there was no shortage of oxygen but the spread of the fire was limited

  20. Mission Critical Occupation (MCO) Charts

    Data.gov (United States)

    Office of Personnel Management — Agencies report resource data and targets for government-wide mission critical occupations and agency specific mission critical and/or high risk occupations. These...

  1. Estimates of power deposited via cesium/barium beta and gamma radiation captured in components of a Hanford cesium chloride capsule and by components of overpacked capsules placed in an interim dry storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Roetman, V.E., Westinghouse Hanford

    1996-12-23

    The deposition of power in Hanford cesium chloride capsules and in the components of design concepts for overpacking and interim storage were determined as requested (Randklev, 1996a). The power deposition results from the selective capture of gamma and beta radiation coming from the decay of the 137CS isotope in the CsCl contained in the capsules. The following three cases were analyzed: (a) a single CsCl capsule, (b) an overpack containing eight CsCl capsules, and (c) an infinite square array of such overpacks as placed in tubes of a interim dry storage facility. The power deposition was expressed as watts per gram for each of the respective physical design components in these three cases. Per the analyses request and guidance (Randklev 1996a), the primary analysis objective was to characterize, for each case, the power deposition across the radial cross-section at the expected axial position of maximum deposition. As requested, this primary part of the analysis work was done using choices for component dimension and material properties that would reasonably characterize the maximum deposition profile across the salt (CsCl) and the inner capsule barrier of the double walled metal capsule system used to construct the Hanford capsules. The secondary objective was to further evaluate the deposition behavior relative to the influence of axial position. The guidance (Randklev 1996a) also requested 1797 an analysis case that involved a lag-storage pit in a hot-cell, in which a cylindrical metal basket from a transportation cask would be used to position several capsules in the lag-storage pit. Although the basic model for the lag storage concept evaluation was essentially completed by the end of FY-96, the analysis was not run because of the need to prioritize and limit the work scope due to funding limitations for FY-97. The specific purpose for performing the subject set of analyses (Randklev 1996a) is to obtain power deposition values (i.e., per the decay of T37cs

  2. Development of a constitutive model for the plastic deformation and creep of copper and its use in the estimate of the creep life of the copper canister

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Kjell [Matsafe AB, Stockholm (Sweden)

    2006-12-15

    A previously developed model for the plastic deformation and creep of copper (included as an Appendix to the present report) has been used as the basis for a discussion on the possibility of brittle creep fracture of the copper canister during long term storage of nuclear waste. Reported creep tests on oxygen free (OF) copper have demonstrated that copper can have an extremely low creep ductility. However with the addition of about 50 ppm phosphorus to the copper it appears as if the creep brittleness problem is avoided and that type of copper (OFP) has consequently been chosen as the canister material. It is shown in the report that the experiments performed on OFP copper does not exclude the possibility of creep brittleness of OFP copper in the very long term. The plasticity and creep model has been used to estimate creep life under conditions of intergranular creep cracking according to a model formulated by Cocks and Ashby. The estimated life times widely exceed the design life of the canister. However the observations of creep brittleness in OF copper indicate that the Cocks-Ashby model probably does not apply to the OF copper. Thus additional calculations have been done with the plasticity and creep model in order to estimate stress as a function of time for the probably most severe loading case of the canister with regard to creep failure, an earth quake shear. Despite the fact that the stress in the canister will remain at the 100 MPa level for thousands of years after an earth quake the low temperature, about 50 deg C or less, will make the solid state diffusion process assumed to control the brittle cracking process, too slow to lead to any significant brittle creep cracking in the canister.

  3. Inspection of copper canister for spent nuclear fuel by means of ultrasound. Copper characterization, FSW monitoring with acoustic emission and ultrasonic imaging

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Engholm, Marcus; Olofsson, Tomas (Uppsala Univ., Signals and Systems, Dept. of Technical Sciences, Uppsala (Sweden))

    2009-08-15

    This report contains the research results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in 2008. The first part of the report is concerned with aspects related to ultrasonic attenuation of copper material used for canisters. We present results of attenuation measurement performed for a number of samples taken from a real canister; two from the lid and four from different parts of canister wall. Ultrasonic attenuation of the material originating from canister lid is relatively low (less that 50 dB/m) and essentially frequency independent in the frequency range up to 5 MHz. However, for the material originating from the extruded canister part considerable variations of the attenuation are observed, which can reach even 200 dB/m at 3.5 MHz. In the second part of the report we present further development of the concept of the friction stir welding process monitoring by means of multiple sensors formed into a uniform circular array (UCA). After a brief introduction into modeling Lamb waves and UCA we focus on array processing techniques that enable estimating direction of arrival of multimodal Lamb waves. We consider two new techniques, the Capon beamformer and the broadband multiple signal classification technique (MUSIC). We present simulation results illustrating their performance. In the final part we present the phase shift migration algorithm for ultrasonic imaging of layered media using synthetic aperture concept. We start from explaining theory of the phase migration concept, which is followed by the results of experiments performed on copper blocks with drilled holes. We show that the proposed algorithm performs well for immersion inspection of metal objects and yields both improved spatial resolution and suppressed grain noise

  4. Oxidative dissolution of spent fuel and release of nuclides from a copper/iron canister. Model developments and applications

    Energy Technology Data Exchange (ETDEWEB)

    Longcheng Liu

    2001-12-01

    Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H{sub 2}, using UO{sub 2} (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H{sub 2} on the fuel surface. It will reprecipitate as UO{sub 2} (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H{sub 2} are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed

  5. Horizontal deposition of canisters for spent nuclear fuel. Summary of the KBS-3H Project 2004-2007

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

    SKB and Posiva both selected the KBS-3 method for the geologic disposal of spent nuclear fuel. The KBS-3 method relies on stable and favourable conditions of the bedrock, long-lived canisters containing the spent fuel and the buffer functions of clay surrounding the canister. The reference design is the KBS-3V, in which the canisters with spent nuclear fuel are emplaced vertically in individual deposition holes. For a number of years SKB and Posiva have also jointly studied a design in which the canisters are instead serially emplaced in long horizontal drifts (KBS-3H). The drivers behind the development of the KBS-3H concept are that both cost and environmental impact could be reduced without compromising long-term safety. There are many similarities between KBS-3H and KBS-3V as both designs are based on the KBS-3 method. The main objectives of KBS-3H Project 2004-2007 were to demonstrate that the deposition alternative is technically feasible and that it fulfils the same long-term safety requirements as KBS-3V. These main objectives have only been partially met owing to the restrictions imposed before the start of the project and during its execution. More work is needed for the full demonstration of the engineering feasibility with due consideration to anticipated, site-specific conditions. In KBS-3H Project 2004-2007, it was demonstrated that it was possible to excavate horizontal drifts that would fulfil most of the stringent requirements on geometry dictated by the use of current standard technology. It was further demonstrated that it is possible to emplace a 46-tonne supercontainer in a deposition drift using water-cushion technology. A critical1 issue for the robustness of the KBS-3H during emplacement and saturation is that the groundwater seepage into the deposition drift is low (< 0.1 l/min over the entire length of the supercontainer section) as higher inflow may cause piping/erosion of the buffer during the saturation period. A Mega-Packer was

  6. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems` Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment.

  7. A FRAMEWORK TO DEVELOP FLAW ACCEPTANCE CRITERIA FOR STRUCTURAL INTEGRITY ASSESSMENT OF MULTIPURPOSE CANISTERS FOR EXTENDED STORAGE OF USED NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P.; Sindelar, R.; Duncan, A.; Adams, T.

    2014-04-07

    A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

  8. 新型舰载同心筒发射过程流场研究%Launching Process of the New Type Shipborne Concentric Canister Launcher

    Institute of Scientific and Technical Information of China (English)

    邵立武; 姜毅; 马艳丽; 王伟臣

    2011-01-01

    The internal-external canister space should be enough to exhaust jet flow when the concentric canister launcher is used. The common-frame launch is used for the shipborne weapon which would cause the small sizes of the internal canister and ballistic missile. To solve the problem, the external canister is designed to be rectangular. The internal-external canister space increases with the same common frame and the missile temperature decreases. The three-dimensional dynamic meshes were used to study the launching process of the concentric canister launcher. The zone moving and dynamic methods were used to update the meshes. The results showed that the location curve accord well with experiment. The missile surface temperature decreased greatly with the new type concentric canister launcher.%采用同心筒垂直发射装置,内外筒要保证足够的间隙尺寸用来排导燃气,目前舰载发射均采用通垂方式,势必使得内筒尺寸较小,从而导弹的直径也就较小,不利于发挥弹道导弹的优势.在此基础上提出将传统的外筒设计为一方形结构,在与相同尺寸的通垂架相连接的前提下,增加了内外筒之间燃气排导空间,降低发射过程中导弹表面的温度.计算中使用三维动网格更新方法对同心筒发射过程进行了计算研究,网格更新方法采用域动分层法.结果表明,导弹运动位移曲线与试验符合较好,新型同心筒方案使得整个发射过程中导弹表面的温度均明显降低.

  9. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 1 - FSW

    Energy Technology Data Exchange (ETDEWEB)

    Purhonen, T.

    2014-05-15

    The purpose of this report is to gather together comprehensive information concerning FSW as an optional welding method for welding the nuclear waste copper canister at the disposal facility. This report discusses the current situation, knowledge of the process and information concerning results of the development and research work related to welding thick copper and the special needs of the disposal environment. Most of the research work and development work has been done by Posiva's Swedish partner SKB, Swedish Nuclear Fuel and Waste Management Co. SKB chose FSW as their reference welding method in 2005. FSW (friction stir welding) is a solid-state welding method, invented in 1991, in which frictional heat is generated between the tool and the weld metal, causing the metal to soften, normally without reaching the melting point, and allowing the tool to traverse the joint line. Friction stir welding can be used for joining many types of materials and material combinations, if the tool materials and designs can be found which operate at the forging temperature of the workpiece. The general requirements for the copper canister weld and base material are presented in Posiva's VAHA-system, which sets the most critical values or demands concerning the short- and long-term properties or other needs. The sections in this report are set out in a similar way as in the VAHA-system. Concerning the results from the research and development work, it can be said that FS weld material fulfils the values set by VAHA. The quality of the welds fulfils the set demands for intact weld material and the welding process is robust using an automatic control system. There still remains work concerning the acceptance procedure for the welding process and other open issues which are described in this report. (orig.)

  10. X-38: Parachute Canister Fired from Plywood Mockup during Flight Termination System Test

    Science.gov (United States)

    1996-01-01

    The canister containing a seven-foot-diameter X-38 Flight Termination System (FTS) parachute is launched safely away from a plywood mockup of the X-38 by a pyrotechnic firing system on December 19, 1996, at NASA Dryden Flight Research Center, Edwards, California. The test was economically accomplished by mounting the mockup of the X-38's aft end, minus vertical stabilizers, on a truck prior to installation in the X-38. The X-38 Crew Return Vehicle (CRV) research project is designed to develop the technology for a prototype emergency crew return vehicle, or lifeboat, for the International Space Station. The project is also intended to develop a crew return vehicle design that could be modified for other uses, such as a joint U.S. and international human spacecraft that could be launched on the French Ariane-5 Booster. The X-38 project is using available technology and off-the-shelf equipment to significantly decrease development costs. Original estimates to develop a capsule-type crew return vehicle were estimated at more than $2 billion. X-38 project officials have estimated that development costs for the X-38 concept will be approximately one quarter of the original estimate. Off-the-shelf technology is not necessarily 'old' technology. Many of the technologies being used in the X-38 project have never before been applied to a human-flight spacecraft. For example, the X-38 flight computer is commercial equipment currently used in aircraft and the flight software operating system is a commercial system already in use in many aerospace applications. The video equipment for the X-38 is existing equipment, some of which has already flown on the space shuttle for previous NASA experiments. The X-38's primary navigational equipment, the Inertial Navigation System/Global Positioning System, is a unit already in use on Navy fighters. The X-38 electromechanical actuators come from previous joint NASA, U.S. Air Force, and U.S. Navy research and development projects. Finally

  11. 湿式独立自排导垂直发射技术研究%Research on the Wet-type Concentric Canister Launcher

    Institute of Scientific and Technical Information of China (English)

    马艳丽; 姜毅; 王伟臣; 刘伯伟; 颜凤

    2011-01-01

    To study the overhigh pressure in the canister during the wet-type canister missile vertical launching, the wet-type concentric canister launchers of various structure parameters are calculated numerically. The launching process of the concentric canister with 17mm space between the inside and outside and single-canister are analyzed numerically with the dynamic update method. Considering the effect of water vaporization, the flow field of vapor and liquid is resolved by Mixture model of dual-phase flow. The zone moving and dynamic laying method was adopted in the meshes update. The result indicates that the smaller space between the inside and outside or the distance between nozzle and rear cover, the bigger pressure in the canister, and the guiding cone in the bottom can obviously decrease the pressure on the rear cover. Therefore, the calculation result accords with the experiment well.%湿式独立自排导垂直发射技术在发射过程中可能会存在发射筒内压力过高的问题,就不同结构参数的湿式独立自排导垂直发射装置进行数值计算,并采用动网格对内外筒间距为17 mm及单筒发射的导弹发射过程进行数值研究和分析.计算中考虑水的汽化效应,采用Mixture两相流计算模型求解气液两相流场,网格更新方法采用域动分层法.结果表明,内外筒间距越小,燃气的排导越受限制,筒内的压力越大;喷管出口与后盖部的距离越小,筒内压力越大;底部加导流锥对于降低后盖上的压力作用明显,计算结果与试验符合较好.

  12. B类滤毒罐防护磷化氢性能评价探讨%Evaluation of B-Type Canister Protective Performance against Phosphine

    Institute of Scientific and Technical Information of China (English)

    汪东旺; 李泽; 赵鑫华; 尹维东; 李志坚; 元以栋

    2012-01-01

    磷化氢是粮食仓储企业使用效果最好的杀虫剂,是一种剧毒的气体熏蒸剂。所以对于从业人员来讲,安全防护就显得格外重要。长期以来粮食仓储企业一直使用自吸过滤式防毒面具(配套选用B类滤毒罐)为首选器材。近期业内对B类滤毒罐防护磷化氢的有效性提出质疑,为此,本文通过性能评价试验,说明B类滤毒罐防护磷化氢是有效的,并从理论上说明B类滤毒罐防护磷化氢是有科学依据的。%Phosphine is the best pesticides of the grain storage enterprise, and is the toxic gaseous fumigant. For users, it is important for the security. For a long time, the grain storage enterprise uses the serf-absorption filtering gas mask (selecting the B-type canister) . Recently, the protective performance against phosphine of B-type canister is doubted, through the test, this paper will explain the effective and scientific theory of B-type canister.

  13. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Phased arrays, ultrasonic imaging and nonlinear acoustics

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Ping Wu; Wennerstroem, Erik [Uppsala Univ. (Sweden). Signals and Systems

    2004-09-01

    This report contains the research results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in years 2003/2004. After a short introduction a review of beam forming fundamentals required for proper understanding phased array operation is included. The factors that determine lateral resolution during ultrasonic imaging of flaws in solids are analyzed and results of simulations modelling contact inspection of copper are presented. In the second chapter an improved synthetic aperture imaging (SAI) technique is introduced. The proposed SAI technique is characterized by an enhanced lateral resolution compared with the previously proposed extended synthetic aperture focusing technique (ESAFT). The enhancement of imaging performance is achieved due to more realistic assumption concerning the probability density function of scatterers in the region of interest. The proposed technique takes the form of a two-step algorithm using the result obtained in the first step as a prior for the second step. Final chapter contains summary of our recent experimental and theoretical research on nonlinear ultrasonics of unbounded interfaces. A new theoretical model for rough interfaces is developed, and the experimental results from the copper specimens that mimic contact cracks of different types are presented. Derivation of the theory and selected measurement results are given in appendix.

  14. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Nonlinear acoustics, synthetic aperture imaging

    Energy Technology Data Exchange (ETDEWEB)

    Lingvall, Fredrik; Ping Wu; Stepinski, Tadeusz [Uppsala Univ., (Sweden). Dept. of Materials Science

    2003-03-01

    This report contains results concerning inspection of copper canisters for spent nuclear fuel by means of ultrasound obtained at Signals and Systems, Uppsala University in year 2001/2002. The first chapter presents results of an investigation of a new method for synthetic aperture imaging. The new method presented here takes the form of a 2D filter based on minimum mean squared error (MMSE) criteria. The filter, which varies with the target position in two dimensions includes information about spatial impulse response (SIR) of the imaging system. Spatial resolution of the MMSE method is investigated and compared experimentally to that of the classical SAFT and phased array imaging. It is shown that the resolution of the MMSE algorithm, evaluated for imaging immersed copper specimen is superior to that observed for the two above-mentioned methods. Extended experimental and theoretical research concerning the potential of nonlinear waves and material harmonic imaging is presented in the second chapter. An experimental work is presented that was conducted using the RITEC RAM-5000 ultrasonic system capable of providing a high power tone-burst output. A new method for simulation of nonlinear acoustic waves that is a combination of the angular spectrum approach and the Burger's equation is also presented. This method was used for simulating nonlinear elastic waves radiated by the annular transducer that was used in the experiments.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-01

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

  16. Volatile Profiles of Emissions from Different Activities Analyzed Using Canister Samplers and Gas Chromatography-Mass Spectrometry (GC/MS) Analysis: A Case Study

    Science.gov (United States)

    Orecchio, Santino; Fiore, Michele; Barreca, Salvatore; Vara, Gabriele

    2017-01-01

    The objective of present study was to identify volatile organic compounds (VOCs) emitted from several sources (fuels, traffic, landfills, coffee roasting, a street-food laboratory, building work, indoor use of incense and candles, a dental laboratory, etc.) located in Palermo (Italy) by using canister autosamplers and gas chromatography-mass spectrometry (GC-MS) technique. In this study, 181 VOCs were monitored. In the atmosphere of Palermo city, propane, butane, isopentane, methyl pentane, hexane, benzene, toluene, meta- and para-xylene, 1,2,4 trimethyl benzene, 1,3,5 trimethyl benzene, ethylbenzene, 4 ethyl toluene and heptane were identified and quantified in all sampling sites. PMID:28212294

  17. High Level ab initio Predictions of the Energetics of mCO2•(H2O)n (n = 1-3, m = 1-12) Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Thanthiriwatte, Sahan; Duke, Jessica R.; Jackson, Virgil E.; Felmy, Andrew R.; Dixon, David A.

    2012-10-04

    Electronic structure calculations at the correlated molecular orbital theory and density functional theory levels have been used to generate a reliable set of clustering energies for up to three water molecules in carbon dioxide clusters up to n = 12. The structures and energetics are dominated by Lewis acid-base interactions with hydrogen bonding interactions playing a lesser energetic role. The actual binding energies are somewhat larger than might be expected. The correlated molecular orbital MP2 method and density functional theory with the ωB97X exchange-correlation functional provide good results for the energetics of the clusters but the B3LYP and ωB97X-D functionals do not. Seven CO2 molecules form the first solvent shell about a single H2O with four CO2 molecules interacting with the H2O via Lewis acid-base interactions, two CO2 interacting with the H2O by hydrogen bonds, and the seventh CO2 completing the shell. The Lewis acid-base and weak hydrogen bond interactions between the water molecules and the CO2 molecules are strong enough to disrupt the trimer ring configuration for as few as seven CO2 molecules. Calculated 13C NMR chemical shifts for mCO2•(H2O)n show little change with respect to the number of H2O or CO2 molecules in the cluster. The O-H stretching frequencies do exhibit shifts that can provide information about the interactions between water and CO2 molecules.

  18. High-level ab initio predictions of the energetics of mCO2·(H2O)n (n = 1-3, m = 1-12) clusters.

    Science.gov (United States)

    Thanthiriwatte, K Sahan; Duke, Jessica R; Jackson, Virgil E; Felmy, Andrew R; Dixon, David A

    2012-10-04

    Electronic structure calculations at the correlated molecular orbital theory and density functional theory levels have been used to generate a reliable set of clustering energies for up to three water molecules in carbon dioxide clusters up to n = 12. The structures and energetics are dominated by Lewis acid-base interactions with hydrogen-bonding interactions playing a lesser energetic role. The actual binding energies are somewhat larger than might be expected. The correlated molecular orbital MP2 method and density functional theory with the ωB97X exchange-correlation functional provide good results for the energetics of the clusters, but the B3LYP and ωB97X-D functionals do not. Seven CO(2) molecules form the first solvent shell about a single H(2)O with four CO(2) molecules interacting with the H(2)O via Lewis acid-base interactions, two CO(2) interacting with the H(2)O by hydrogen bonds, and the seventh CO(2) completing the shell. The Lewis acid-base and weak hydrogen bond interactions between the water molecules and the CO(2) molecules are strong enough to disrupt the trimer ring configuration for as few as seven CO(2) molecules. Calculated (13)C NMR chemical shifts for mCO(2)·(H(2)O)(n) show little change with respect to the number of H(2)O or CO(2) molecules in the cluster. The O-H stretching frequencies do exhibit shifts that can provide information about the interactions between water and CO(2) molecules.

  19. Long-term integrity of copper overpack

    Energy Technology Data Exchange (ETDEWEB)

    Holmstroem, Stefan; Salonen, Jorma; Auerkari, Pertti (VTT, Esbo (FI)); Saukkonen, Tapio (Helsinki Univ. of Technology, Esbo (FI))

    2007-05-15

    The results from extended uniaxial and multiaxial creep testing confirm the earlier indications of microstructural changes at relatively low temperatures (150-175 deg C) in Cu-OFP. These changes are probably related to recovery processes directed by the favourable crystallographic orientation on one side the related grain boundary, resulting in characteristically widening grain boundary zones. With further straining, these zones become chains of small grains decorating the original grain boundaries. The observed microstructural changes do not appear to represent particular disadvantages in terms of remaining life. In creep testing with natural weld defects (FSW, inclusion sheet 20% of cross-section), the results show much faster decreasing creep strength in time than what is observed for base material or welds without defects. However, extrapolation to 50 MPa stress level across such a defective region would still suggest a safe life of approximately 26,000 years in spite of much elevated testing temperature (175 deg C) from expected service temperature (below 100 deg C). For predicting mechanical behaviour, a creep model has been developed to include the full creep curves in a simple and robust manner. The model has been adapted to the most recent creep testing results (up to about 48,000 h in uniaxial testing). Applying this model for the extrapolated case of steady loading at 100 deg C / 50 MPa predicts time to 10% strain of about one million years. For comparison on creep ductility, also a testing program on low-phosphorus (OFHC) copper was initiated. The testing program with model vessels was completed after confirming safe short term limit load predictions. This program continues with compact tension specimens to study the potential combined effect of creep and corrosion in simulated groundwater

  20. HYDRA-I: a three-dimensional finite difference code for calculating the thermohydraulic performance of a fuel assembly contained within a canister

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.

    1980-12-01

    A finite difference computer code, named HYDRA-I, has been developed to simulate the three-dimensional performance of a spent fuel assembly contained within a cylindrical canister. The code accounts for the coupled heat transfer modes of conduction, convection, and radiation and permits spatially varying boundary conditions, thermophysical properties, and power generation rates. This document is intended as a manual for potential users of HYDRA-I. A brief discussion of the governing equations, the solution technique, and a detailed description of how to set up and execute a problem are presented. HYDRA-I is designed for operation on a CDC 7600 computer. An appendix is included that summarizes approximately two dozen different cases that have been examined. The cases encompass variations in fuel assembly and canister configurations, power generation rates, filler materials, and gases. The results presented show maximum and various local temperatures and heat fluxes illustrating the changing importance of the three heat transfer modes. Finally, the need for comparison with experimental data is emphasized as an aid in code verification although the limited data available indicate excellent agreement.

  1. Whole air canister sampling coupled with preconcentration GC/MS analysis of part-per-trillion levels of trimethylsilanol in semiconductor cleanroom air.

    Science.gov (United States)

    Herrington, Jason S

    2013-08-20

    The costly damage airborne trimethylsilanol (TMS) exacts on optics in the semiconductor industry has resulted in the demand for accurate and reliable methods for measuring TMS at trace levels (i.e., parts per trillion, volume per volume of air [ppt(v)] [~ng/m(3)]). In this study I developed a whole air canister-based approach for field sampling trimethylsilanol in air, as well as a preconcentration gas chromatography/mass spectrometry laboratory method for analysis. The results demonstrate clean canister blanks (0.06 ppt(v) [0.24 ng/m(3)], which is below the detection limit), excellent linearity (a calibration relative response factor relative standard deviation [RSD] of 9.8%) over a wide dynamic mass range (1-100 ppt(v)), recovery/accuracy of 93%, a low selected ion monitoring method detection limit of 0.12 ppt(v) (0.48 ng/m(3)), replicate precision of 6.8% RSD, and stability (84% recovery) out to four days of storage at room temperature. Samples collected at two silicon wafer fabrication facilities ranged from 10.0 to 9120 ppt(v) TMS and appear to be associated with the use of hexamethyldisilazane priming agent. This method will enable semiconductor cleanroom managers to monitor and control for trace levels of trimethylsilanol.

  2. Cross-section measurement for Ni(n,x)58(m+g)Co,Ni(n,x)60mCo,Ni(n,x)61Co and Ni(n,x)62mCo reactions induced by neutrons around 14MeV

    Institute of Scientific and Technical Information of China (English)

    FANG Kai-Hong; XU Xiao-San; LAN Chang-Lin; UAN Ji-Long; KONG Xiang-Zhong

    2008-01-01

    The cross sections of Ni(n,x)58(m+g)CO,Ni(n,x)80mCo,Ni(n,x)61Co and Ni(n,x)62mCo reactions induced by neutrons around [14]MeV were measured in this work and calculated by a previously developed formula in this work.The neutron flux was determined using the monitor reaction 27Al(n,α)24Na and the neutron energies were measured with the method of cross-section ratios for 89Zr(n,2n)89Zr to 93Nb(n,2n)92mNbreactions.

  3. Engineered Barrier System - Assessment of the Corrosion Properties of Copper Canisters. Report from a Workshop. Synthesis and extended abstract

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Peter (ed.) [Quintessa Ltd., Henley-on-Thames (GB)] (and others)

    2006-03-15

    assumption turns out not to be valid at some stage during the repository evolution. Workshop participants suggested a need for SKI to review SKB's canister corrosion model in more detail as part of future safety assessment reviews (calculations, assumptions and data). Additional experimental work might be needed for the assessment of copper corrosion in high chloride environments and with simultaneous presence of chloride and sulphide. It is essential that altogether consistent facts, understanding and models are used when developing an argument. Any inconsistency regarding these three aspects (facts, understanding, models) needs to be identified. An example would be if thermodynamic data and theoretical calculations suggest that corrosion will not happen, while kinetic data (experimental results) suggest a significant corrosion rate. For future safety assessments, SKB is recommended to use a consistent template for the handling of different corrosion mechanisms even if their final treatment will be quite different. This may include e.g. an extended application of the exclusion principle and/or application of the decision tree approach (as applied for stress corrosion cracking in the Canadian programme). However, it should be noted that the reliability of the exclusion principle depends on the quantity and quality of information on which it is based, and that more explicit criteria might be needed to support the decision tree approach. There is also a need for a well structured approach to handling uncertainties. Examples include those that can be characterised as variability (welding defects, sulphide content of groundwater and bentonite) and as lack of knowledge (e.g. microbial viability, the existence of an unidentified groundwater component affecting corrosion or an unknown corrosion mechanism). A suitable combination of a probabilistic application of the main copper corrosion model, well supported calculation cases with mechanistic models and possibly a selection

  4. Engineered Barrier System - Assessment of the Corrosion Properties of Copper Canisters. Report from a Workshop. Synthesis and extended abstract

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Peter (ed.) [Quintessa Ltd., Henley-on-Thames (GB)] (and others)

    2006-03-15

    assumption turns out not to be valid at some stage during the repository evolution. Workshop participants suggested a need for SKI to review SKB's canister corrosion model in more detail as part of future safety assessment reviews (calculations, assumptions and data). Additional experimental work might be needed for the assessment of copper corrosion in high chloride environments and with simultaneous presence of chloride and sulphide. It is essential that altogether consistent facts, understanding and models are used when developing an argument. Any inconsistency regarding these three aspects (facts, understanding, models) needs to be identified. An example would be if thermodynamic data and theoretical calculations suggest that corrosion will not happen, while kinetic data (experimental results) suggest a significant corrosion rate. For future safety assessments, SKB is recommended to use a consistent template for the handling of different corrosion mechanisms even if their final treatment will be quite different. This may include e.g. an extended application of the exclusion principle and/or application of the decision tree approach (as applied for stress corrosion cracking in the Canadian programme). However, it should be noted that the reliability of the exclusion principle depends on the quantity and quality of information on which it is based, and that more explicit criteria might be needed to support the decision tree approach. There is also a need for a well structured approach to handling uncertainties. Examples include those that can be characterised as variability (welding defects, sulphide content of groundwater and bentonite) and as lack of knowledge (e.g. microbial viability, the existence of an unidentified groundwater component affecting corrosion or an unknown corrosion mechanism). A suitable combination of a probabilistic application of the main copper corrosion model, well supported calculation cases with mechanistic models and possibly a selection

  5. Evaluating the use of PAO (4 cSt polyalphaoelfin) oil instead of DOP (di-octyl phthalate) oil for measuring the aerosol capture of nuclear canister filters

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-07-18

    This document details the distinction between using PAO (4 cSt polyalphaoelfin) oil instead of DOP (di-octyl phthalate) oil for measuring the aerosol capture of filters. This document is developed to justify the use of PAO rather than DOP for evaluating the performance of filters in the SAVY 4000 and Hagan containers. The design criteria (Anderson et al, 2012) for purchasing SAVY 4000 containers and the Safety Analysis Report for the SAVY 4000 Container Series specified that the filter must “capture greater than 99.97% of 0.45 μm mean diameter dioctyl phthalate (DOP) aerosol at the rated flow with a DOP concentration of 65±15 micrograms per liter.”This corresponds to a leakage percent of 0.03% (3.0x10-2). The density of DOP oil is 985 kg/m3 and the density of PAO oil is 819 kg/m3. ATI Test Inc measured the mass mean diameter of aerosol distributions produced by a single Laskin type III-A nozzle operating at a 20 psig air pressure as 0.563 μm for DOP oil and 0.549 μm for PAO oil. (See Appendix A.) For both types of oil in this document, the single fiber method calculated the leakage percent to be 4.4x10-5 for DOP oil and 4.7x10-5 for PAO oil. Although the percent error between these two quantities is 7.7%, these calculated leakage percent values are more than two orders of magnitude less than the criterion specified in the SAVY canister SAR. As a point of reference, the photometer used to measure the SAVY canister filter performance cannot resolve values for the leakage percent below 1.0x10-5. Additionally, over a range of particle sizes from 0.01 μm to 3.0 μm, there was less than 4.0x10-5 error between the calculated filter efficiency for the two types of oil at any particular particle size diameter. In conclusion, the difference between using DOP and PAO for testing SAVY canister filters is of inconsequential concern.

  6. Stack Flow Rate Changes and the ANSI/N13.1-1999 Qualification Criteria: Application to the Hanford Canister Storage Building Stack

    Energy Technology Data Exchange (ETDEWEB)

    Flaherty, Julia E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Glissmeyer, John A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-02-29

    The Canister Storage Building (CSB), located in the 200-East Area of the Hanford Site, is a 42,000 square foot facility used to store spent nuclear fuel from past activities at the Hanford Site. Because the facility has the potential to emit radionuclides into the environment, its ventilation exhaust stack has been equipped with an air monitoring system. Subpart H of the National Emissions Standards for Hazardous Air Pollutants requires that a sampling probe be located in the exhaust stack in accordance with criteria established by the American National Standards Institute/Health Physics Society Standard N13.1-1999, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities.

  7. Geological Disposal of Nuclear Waste: Investigating the Thermo-Hygro-Mechanical-Chemical (THMC) Coupled Processes at the Waste Canister- Bentonite Barrier Interface

    Science.gov (United States)

    Davies, C. W.; Davie, D. C.; Charles, D. A.

    2015-12-01

    Geological disposal of nuclear waste is being increasingly considered to deal with the growing volume of waste resulting from the nuclear legacy of numerous nations. Within the UK there is 650,000 cubic meters of waste safely stored and managed in near-surface interim facilities but with no conclusive permanent disposal route. A Geological Disposal Facility with incorporated Engineered Barrier Systems are currently being considered as a permanent waste management solution (Fig.1). This research focuses on the EBS bentonite buffer/waste canister interface, and experimentally replicates key environmental phases that would occur after canister emplacement. This progresses understanding of the temporal evolution of the EBS and the associated impact on its engineering, mineralogical and physicochemical state and considers any consequences for the EBS safety functions of containment and isolation. Correlation of engineering properties to the physicochemical state is the focus of this research. Changes to geotechnical properties such as Atterberg limits, swelling pressure and swelling kinetics are measured after laboratory exposure to THMC variables from interface and batch experiments. Factors affecting the barrier, post closure, include corrosion product interaction, precipitation of silica, near-field chemical environment, groundwater salinity and temperature. Results show that increasing groundwater salinity has a direct impact on the buffer, reducing swelling capacity and plasticity index by up to 80%. Similarly, thermal loading reduces swelling capacity by 23% and plasticity index by 5%. Bentonite/steel interaction studies show corrosion precipitates diffusing into compacted bentonite up to 3mm from the interface over a 4 month exposure (increasing with temperature), with reduction in swelling capacity in the affected zone, probably due to the development of poorly crystalline iron oxides. These results indicate that groundwater conditions, temperature and corrosion

  8. State of Washington Department of Health radioactive air emission notice of construction phase 1 for spent nuclear fuel project - hot conditioning system annex, project W-484

    Energy Technology Data Exchange (ETDEWEB)

    Turnbaugh, J.E.

    1996-08-15

    This notice of construction (NOC) provides information regarding the source and the estimated annual possession quantity resulting from the operation of the Hot Conditioning System Annex (HCSA). This information will be discussed again in the Phase II NOC, providing additional details on emissions generated by the operation of the HCSA. This Phase I NOC is defined as construct in the substructure, including but limited to, pouring the concrete for the floor; construction of the process pits and exterior walls; making necessary interface connections to the Canister Storage Building (CSB) ventilation and utility systems for personnel comfort; and extending the multi-canister over-pack (MCO) handling machine rails into the HCSA. A Phase II NOC will be submitted for approval prior to installation and is defined as the completion of the HCSA, which will consist of installation of Hot Conditioning System Equipment (HCSA), air emissions control equipment, and emission monitoring equipment. About 80 percent of the U.S. Department of Energy`s spent nuclear fuel (SNF) inventory is stored under water in the Hanford Site K Basins. Spent nuclear fuel in the K West Basin is contained in closed canisters, while the SNF in the K East Basin is contained in open canisters, which allow free release of corrosion products to the K East Basin water. Storage in the K Basins was originally intended to be on an as-needed basis to sustain operation of the N Reactor while the Plutonium-Uranium Extraction (PUREX) Plant was refurbished and restarted. The decision in December 1992 to deactivate the PUREX Plant left approximately 2,300 MT (2,530 tons) of N Reactor SNF in the K Basins with no means for near-term removal and processing. The HCSA will be constructed as an addition to the CSB and will contain the HCSA. The hot conditioning system (HCS) will remove chemically-bound water and will passivate the exposed uranium surfaces associated,with the SNF. The HCSA will house seven hot

  9. Inspection of copper canisters for spent nuclear fuel by means of ultrasonic array system. Modelling, defect detection and grain noise estimation

    Energy Technology Data Exchange (ETDEWEB)

    Wu Ping; Stepinski, T. [Uppsala Univ., (Sweden). Dept. of Material Science

    1998-07-01

    The work presented in the report has been split into three overlapping tasks which have the following objectives: (1) development of beam-forming tools, and verification of modeling tools; (2) investigation of detection and resolution limits; (3) evaluation of attenuation, estimation and suppression of grain noise. For beam-forming tools, a method of designing steered and/or focused beams in immersed solids is presented based on geometrical acoustics. Presently, the beam designs are only related to delays but not to apodization. These focused, steered beams are intended to be used for sizing defects and inspecting the regions close to canisters outer walls. The modeling tool developed previously for simulating elastic fields radiated by planar arrays into immersed solids has been verified by comparing with the results obtained from PASS, a software developed by Dr. Didier Cassereau, France. The results from our modeling tool are in excellent agreement with those from PASS. Since the array coming with the ALLIN ultrasonic array system is not planar, but cylindrically curved in elevation, and it works not in transmission mode, but in pulse echo mode, the above modeling tool for the planar arrays cannot be applied directly. Therefore, the modeling tool has been upgraded for the ALLIN array. The theory underlying this modeling tool is the extended angular spectrum approach (ASA) which was developed based on the conventional ASA that only applies to planar sources. Experimental verification of the modeling tool has shown that the results from the tool agree very well with the measurements. To quantify the fields from the ALLIN array and to facilitate the comparison of simulated results with the measured ones, the ALLIN array system has been calibrated based on the existing functionality, and an analytical model has been proposed for simulating measured acoustic echo pulses. To investigate the detection and resolution limits, we have carried out a series of experiments

  10. Interior Ballistic Modeling and Simulation of Underwater Launched Missile Using Concentric Canister Launcher%同心筒水下发射内弹道建模与仿真研究

    Institute of Scientific and Technical Information of China (English)

    袁绪龙; 王亚东; 刘维

    2013-01-01

    To construct a fast calculation method of interior ballistics of underwater launched missile using Concentric Canister Launcher(CCL),a simulation model of CCL was established according to the first law of thermodynamics,and power characteristics and underwater environment were considered.The empirical parameters in this model were decided using CFD solutions,and they were verified by more calculations.The influences of design parameters on the interior ballistics were studied by the verified model.The simulation shows that the sizes of canister top and bottom restricting parts can be used to adjust launching velocity of missile;the size of canister bottom restricting part can be adapted to modify the acceleration of missile.It can increase the adjusting range of velocity to increase initial volumes of inner and outer canister.The simulation method and results offer reference for engineers.%为构建一种快速的同心筒水下发射内弹道算法,采用热力学第一定律,结合导弹动力装置特性及水下环境需求,建立了同心筒水下发射内弹道计算模型,用CFD结果辨识并校验模型经验参数.应用校验的模型研究了发射装置设计参数对内弹道参数的影响规律.结果表明:筒口、筒底限流尺寸均可用于调速,筒底限流尺寸可用于调节过载,内、外筒初始容积增大可增大调速范围.仿真方法和结果可供工程设计人员参考.

  11. 罐采样与GC-MS联用测定空气中三甲胺%Determining Trimethylamine in Air by Canister Sampling and GC-MS

    Institute of Scientific and Technical Information of China (English)

    黄旭锋

    2015-01-01

    采用真空罐采样,三级冷阱预浓缩仪处理样品,气相色谱-质谱法测定空气中的三甲胺.结果表明,检出限为0.005 mg/m3,相对标准偏差为4.1%,加标回收率为92.0%~108%,能达到国家标准要求.该方法操作简单,测定准确可靠,可用于环境空气中三甲胺的测定.%With Sampling by vacuum canister and processing samples by three stage cold trap,we determined trimethylamine in the air samples by gas chromatography⁃mass spectrometry. The results show that the detection limit was 0.005 mg/m3,the relative standard deviation is 4.1%,and the recovery rate was 92.0%~108%,which can meet the requirements of Emission Standards for Odor Pollutants. This method is simple,accurate and reliable,which can be used for determination of trimethylamine in ambient air.

  12. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. NDE of friction stir welds, nonlinear acoustics, ultrasonic imaging

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Lingvall, Fredrik; Wennerstroem, Erik; Ping Wu [Uppsala Univ., Dept. of Materials Science (Sweden). Signals and Systems

    2004-01-01

    This report contains results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in years 2002/2003. After a short introduction a review of the NDE techniques that have been applied to the assessment of friction stir welds (FSW) is presented. The review is based on the results reported by the specialists from the USA, mostly from the aerospace industry. A separate chapter is devoted to the extended experimental and theoretical research concerning potential of nonlinear waves in NDE applications. Further studies concerning nonlinear propagation of acoustic and elastic waves (classical nonlinearity) are reported. Also a preliminary investigation of the nonlinear ultrasonic detection of contacts and interfaces (non-classical nonlinearity) is included. Report on the continuation of previous work concerning computer simulation of nonlinear propagations of ultrasonic beams in water and in immersed solids is also presented. Finally, results of an investigation concerning a new method of synthetic aperture imaging (SAI) and its comparison to the traditional phased array (PA) imaging and to the synthetic aperture focusing technique (SAFT) are presented. A new spatial-temporal filtering method is presented that is a generalization of the previously proposed filter. Spatial resolution of the proposed method is investigated and compared experimentally to that of classical SAFT and PA imaging. Performance of the proposed method for flat targets is also investigated.

  13. Review of NDE Methods for Detection and Monitoring of Atmospheric SCC in Welded Canisters for the Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sorenson, Ken B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-01-14

    Dry cask storage systems (DCSSs) for used nuclear fuel (UNF) were originally envisioned for storage periods of short duration (~ a few decades). However, uncertainty challenges the opening of a permanent repository for UNF implying that UNF will need to remain in dry storage for much longer durations than originally envisioned (possibly for centuries). Thus, aging degradation of DCSSs becomes an issue that may not have been sufficiently considered in the design phase and that can challenge the efficacy of very long-term storage of UNF. A particular aging degradation concern is atmospheric stress corrosion cracking (SCC) of DCSSs located in marine environments. In this report, several nondestructive (NDE) methods are evaluated with respect to their potential for effective monitoring of atmospheric SCC in welded canisters of DCSSs. Several of the methods are selected for evaluation based on their usage for in-service inspection applications in the nuclear power industry. The technologies considered include bulk ultrasonic techniques, acoustic emission, visual techniques, eddy current, and guided ultrasonic waves.

  14. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Electron beam evaluation, harmonic imaging, materials characterization, and ultrasonic modelling

    Energy Technology Data Exchange (ETDEWEB)

    Wu Ping; Lingvall, Fredrik; Stepinski, Tadeusz [Uppsala Univ. (Sweden). Dept. of Materials Science

    2000-12-01

    This report presents the research in the sixth phase that is concerned with ultrasonic techniques for assessing electron beam (EB) welds in copper canisters. The research has been carried out in three main aspects: (1) comparative inspections of EB welds, (2) EB weld evaluation, and (3) quantitative evaluation of attenuation in copper. Comparative inspections of EB welds in two copper canister blocks have been made by means of ultrasound and radiography. Comparison of the inspected results demonstrate that both techniques complement each other very well. The radiographic technique on the whole gives relatively better spatial resolution but low contrast in radiographs. It can reliably detect voids in EB, but cannot provide information about material structure in the EB weld. Ultrasonic technique provides information about flaw locations and shapes similar to the radiographs. Moreover, it can easily distinguish welded and non-welded zones and be used to study weld's macro- and microstructure. The defects in ultrasonic images often show higher contrast, and some flaw indications may be seen in ultrasonic inspection but not in radiographs. But small flaws are hard to distinguish from grain noise. For EB weld evaluation, first, scattering from EB weld has been investigated using three broadband transducers with different center frequencies. The investigation has shown that more information on scattering and attenuation can be exploited in this case so that the EB welds can be better characterized, and that the best frequency range for characterizing welds is 2 - 5 MHz. Secondly, harmonic imaging (HI) of EB welds have been studied using two different sources of harmonics: (i) transducer harmonics, originating from the high-order resonant modes of transmitters excited by a broadband pulse, and (ii) material harmonics, stemming from the nonlinear distortion of waves propagating in materials. The transducer HI exploits additional information due to transducer harmonics

  15. Inspection of copper canisters for spent nuclear fuel by means of Ultrasonic Array System. Electron beam evaluation, modeling and materials characterization

    Energy Technology Data Exchange (ETDEWEB)

    Ping Wu; Lingvall, F.; Stepinski, T. [Uppsala Univ. (Sweden). Dept. of Material Science

    1999-12-01

    Research conducted in the fifth phase of the SKB's study aimed at developing ultrasonic techniques for assessing EB welds copper canisters is reported here. This report covers three main tasks: evaluation of electron beam (EB) welds, modeling of ultrasonic fields and characterization of copper material. A systematic analysis of ultrasonic interaction and imaging of an EB weld has been performed. From the analysis of histograms of the weld ultrasonic image, it appeared that the porosity tended to be concentrated towards the upper side of a HV weld, and a guideline on how to select the gates for creating C-scans has been proposed. The spatial diversity method (SDM) has shown a limited ability to suppress grain noise both in the parent material (copper) and in the weld so that the ultrasonic image of the weld could be improved. The suppression was achieved at the price of reduced spatial resolution. The ability of wavelet filters to enhance flaw responses has been studied. An FIR (finite impulse response) filter, based on Sombrero mother wavelet, has yield encouraging results concerning clutter suppression. However, the physical explanation for the results is still missing and needs further research. For modeling of ultrasonic fields of the ALLIN array, an approach to computing the SIR (spatial impulse response) of a cylindrically curved, rectangular aperture has been developed. The aperture is split into very narrow strips in the cylindrically curved direction and SIR of the whole aperture by superposing the individual impulse responses of those strips. Using this approach, the SIR of the ALLIN array with a cylindrically curved surface has been calculated. The pulse excitation of normal velocity on the surface of the array, that is required for simulating actual ultrasonic fields, has been determined by measurement in combination with a deconvolution technique. Using the SIR and the pulse excitation obtained, the pulsed-echo fields from the array have been

  16. Development of a method for the study of H{sub 2} gas emission in sealed compartments containing canister copper immersed in O{sub 2}-free water

    Energy Technology Data Exchange (ETDEWEB)

    Bengtsson, Andreas; Chukharkina, Alexandra; Eriksson, Lena; Hallbeck, Bjoern; Hallbeck, Lotta; Johansson, Jessica; Johansson, Linda; Pedersen, Karsten [Microbial Analytics Sweden AB, Moelnlycke (Sweden)

    2013-06-15

    Current models of copper corrosion indicate that copper is not subject to corrosion by water in itself, but that additional components, such as O{sub 2}, chloride or sulphide are needed to initiate a corrosive process. Of late however, a number of reports have suggested that copper may be susceptible to water-induced corrosion in the absence of external constituents affecting the process. The process has been proposed to rely the auto-ionization driven presence of the hydroxide ions in pure water, and to result in the development of atomic hydrogen (H), with subsequent release of H{sub 2} gas. A suggested equilibrium is reached at a partial pressure of H{sub 2} of about 1 mbar (0.1 kPa) in 73 deg C, and the corrosion reaction is proposed to be rate-limited by the supply of hydroxide ions from the water, a process being slower than proposed formation of water from a H{sub 2}-O{sub 2} reaction. In consequence, the presence of O{sub 2} in the system would result in no detectable release of H{sub 2} until all O{sub 2} was consumed, while the absence of O{sub 2} would lead to water-driven corrosion of copper proceeding until the H{sub 2} equilibrium is reached, at a partial H{sub 2} pressure of about 1 mbar. The proposed mechanism presents a novel aspect on copper corrosion processes. By extension, the suggested corrosion process may have implications for proposed strategies for long-term storage of spent nuclear fuel waste (SNF), which in part rely on the long-term (>105 years) integrity of copper canisters stored in anoxic water inundated environments (SKB 2010)

  17. 过滤罐微生物气溶胶过滤效率及其评价方法的研究%A methodological study on testing and evaluating of filtration efficiency of canister against microbial aerosol

    Institute of Scientific and Technical Information of China (English)

    温占波; 赵建军; 李劲松; 王洁; 鹿建春; 李娜

    2009-01-01

    目的 建立防护面具高效过滤罐微生物气溶胶测试评价方法,对过滤罐的实际防护效果进行测试评价.方法 Serratia marcescens作为模式细菌繁殖体气溶胶、Bacillus subtilis var niger芽孢作为模式芽孢气溶胶、噬菌体f2作为模式病毒气溶胶,使用实验室建立的微生物气溶胶检测技术平台,人工发生模式微生物气溶胶,分别在过滤罐过滤前后使用空气微生物采样器进行定量采样,根据过滤前后模式微生物气溶胶的浓度分别计算细菌、芽孢、病毒气溶胶过滤效率.1-1、1-2、1-3、1-44个只含有高效过滤材料的过滤罐分别测试了Serratia marcescens、Bacillus subtilis var niger、噬菌体f2气溶胶的过滤效率.543、544 2个装有活性炭的高效过滤罐测试了对Scrratia marcescens气溶胶的过滤效率.结果 1-1、1-2、1-3 3个高效过滤罐对Serratia marcescens、Bacillus subtilis var niger芽孢、噬菌体f2的气溶胶的过滤效率为100.000%,1-4高效过滤罐对Bacillus subtilis var niger芽孢气溶胶的过滤效率为990997%、Serratia marcescerts和噬菌体f2气溶胶的过滤效率均为100.000%.加入活性炭后543、544 2个过滤罐对Serratia marcescens气溶胶的过滤效率均为100.000%.结论 建立的检测方法可以用于高效过滤罐微生物气溶胶防护效果的评价,高效过滤罐(包括装有活性炭者)微生物气溶胶防护效果均佳.%Objective To establish a testing and evaluating method for filtration efficiency of the eanister against microbial aerosol. Methods Serratia marcescens aerosol served as model of bacterial aerosol, Bacillus subtilis var niger aerosol as model of spores aerosol, bacteriophage f2 aerosol as model of viral aerosol. Employing the microbial aerosol testing platform was established in lab, models of microbial aerosol generated artificially were sampled quantitatively by air samplers before and after filtrating by canisters, respectively. Filtration

  18. 在高湿度环境下用活性炭盒测量氡浓度的研究%STUDY ON RADON CONCENTRATION MONITORING USING ACTIVATED CHARCOAL CANISTERS IN HIGH HUMIDITY ENVIRONMENTS

    Institute of Scientific and Technical Information of China (English)

    王月兴; 王海军; 杨翊方; 秦思昌; 王震涛; 张振江

    2009-01-01

    The effects of humidity on the sensitivity using activated charcoal canisters for measuring radon concentrations in high humidity environments were studied.Every canister filled with 80 g of activated charcoal,and they were exposed to 48 h or 72 h in the relative humidity of 68%,80%,88% and 96% (28℃),respectively.The amount of radon absorbed in the canisters was determined by counting the gamma rays from 214 Pb and 214 Bi (radon progeny).The results showed that counts decreased with the increase of relative humidity.There was a negative linear relationship between count and humidity.In the relative humidity range of 68%-96%,the sensitivity of radon absorption decreased about 2.4% for every 1% (degree) rise in humidity.The results also showed that the exposure time of the activated charcoal canisters should be less than 3 days.%本文研究了在高湿度环境中使用活性炭盒测量氡浓度时湿度对灵敏度的影响.所用的活性炭盒为圆柱形,每一个盒内装80 g活性炭.活性炭盒在相对湿度为68%、80%、88%和96%环境中(28℃)暴露48 h和72 h.在盒内被吸收的氡的量用氡子体214 Ph和214 Bi的γ射线计数确定.实验结果表明,在湿度相同情况下,计数随湿度的增高而降低,两个变量之间呈现负线性相关.在相对湿度68%到96%之间,湿度每增加1%,吸收氡的灵敏度减少约2.4%.在高湿度环境中,活性炭盒的暴露时间不宜超过3天.

  19. Multi-objective optimization for orientator of airdropping launch canister%某运发箱定向器多目标优化设计研究

    Institute of Scientific and Technical Information of China (English)

    胡建国; 仲健林; 马大为; 乐贵高; 周晓和; 杨风波

    2014-01-01

    为实现空投储运发射箱轻量化、保证空投安全性,对定向器进行多目标优化设计。在有限元参数化模型基础上,结合Isight优化平台建立运发箱定向器冲击动力学多目标优化框架;以复合材料定向器铺层厚度、角度为输入变量,定向器质量、最大位移、药柱最大应力为输出变量,采用最优拉丁超立方设计和径向基神经网络方法建立数学近似模型;采用非劣排序遗传算法及模糊集合理论,得到多目标优化模型的Pareto解集及选优方案,并与原方案进行了对比。结果表明:优化方案实现了储运发射箱的轻量化设计,并提高了空投安全性。%To realize the lightweight and guarantee the safety for airdropping launch canister,multi-objective optimization for the orientator was proposed. Firstly,based on the finite element parameterized model,the impact dynamics multi-objective optimization framework was built with Isight platform. Secondly,with layer thickness,layer angle as the input variables and the mass,maximum displacement of orientator,maximum stress of grain as the output variables,the mathematical approximate model was proposed with optimal latin hypercube design and radial basis function neural network method. Finally,on the basis of non-dominated sorting genetic algorithm and fuzzy set theory,the Pareto solutions and the optimum choice were obtained,and compared with the old scheme. The results show that the optimum choice can guarantee the mass of orientator and increase the safety of airdropping.

  20. Waste Package Outer Barrier Stress Due to Thermal Expansion with Various Barrier Gap Sizes

    Energy Technology Data Exchange (ETDEWEB)

    M. M. Lewis

    2001-11-27

    The objective of this activity is to determine the tangential stresses of the outer shell, due to uneven thermal expansion of the inner and outer shells of the current waste package (WP) designs. Based on the results of the calculation ''Waste Package Barrier Stresses Due to Thermal Expansion'', CAL-EBS-ME-000008 (ref. 10), only tangential stresses are considered for this calculation. The tangential stresses are significantly larger than the radial stresses associated with thermal expansion, and at the WP outer surface the radial stresses are equal to zero. The scope of this activity is limited to determining the tangential stresses the waste package outer shell is subject to due to the interference fit, produced by having two different shell coefficients of thermal expansions. The inner shell has a greater coefficient of thermal expansion than the outer shell, producing a pressure between the two shells. This calculation is associated with Waste Package Project. The calculations are performed for the 21-PWR (pressurized water reactor), 44-BWR (boiling water reactor), 24-BWR, 12-PWR Long, 5 DHLW/DOE SNF - Short (defense high-level waste/Department of Energy spent nuclear fuel), 2-MCO/2-DHLW (multi-canister overpack), and Naval SNF Long WP designs. The information provided by the sketches attached to this calculation is that of the potential design for the types of WPs considered in this calculation. This calculation is performed in accordance with the ''Technical Work Plan for: Waste Package Design Description for SR (Ref.7). The calculation is documented, reviewed, and approved in accordance with AP-3.12Q, Calculations (Ref.1).

  1. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L. [Natural Environment Research Council (United Kingdom). British Geological Survey

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  2. Spent nuclear fuel project product specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    1999-02-25

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  3. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-01-20

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  4. Geochemical assessment of nuclear waste isolation. Report of activities during fiscal year 1982

    Energy Technology Data Exchange (ETDEWEB)

    1983-07-01

    The status of the following investigations is reported: canister/overpack-backfill chemical interactions and mechanisms; backfill and near-field host rock chemical interactions mechanisms; far-field host rock geochemical interactions; verification and improvement of predictive algorithms for radionuclide migration; and geologic systems as analogues for long-term radioactive waste isolation.

  5. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L. [Natural Environment Research Council (United Kingdom). British Geological Survey

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  6. 基于有限元动力学的复合材料发射筒多目标优化设计%Multi-objective Optimization of Composites Launch Canister Based on Finite Element Dynamics

    Institute of Scientific and Technical Information of China (English)

    朱孙科; 马大为; 罗天洪; 李士军

    2012-01-01

    To solve the problem of the contact impact between missile and launch canister, we use the implicit scheme in the finite element's (FE) statics and the explicit scheme in FE dynamics to simulate a missile's charging and launching respectively. We use the composites launch canister's ply thickness and its lamination angle as the optimal design variables and define the launch canister's maximum displacement during missile launching and the launch canister's mass as optimization objectives. Taking into account the FE numerical calculation results, we use the Python language to develop the programs and the non-dominant sequencing genetic algorithm II (NSGA-II) which is based on the Pareto strategy to establish the multi-objective optimization model. With the optimization model, we obtain the Pareto front surface curve and the optimal solution. We use the specific stiffness structural ef- ficiency to' compare the performance of the launch canister before and after the optimization ; the comparison results show that optimal design of the launch canister is effective.%针对弹筒接触碰撞问题,分别采用有限元静力学隐式格式和动力学显式格式,模拟了导弹装填及发射过程。并以复合材料发射筒的复合铺层厚度和铺层角度为优化设计变量,定义发射筒在导弹发射过程中筒口产生的最大位移和发射筒质量为优化目标,结合有限元数值计算结果,采用Python语言进行编程,运用基于Pareto策略的改进的非支配排序遗传算法(NSGA-II),建立了多目标优化模型。通过优化求解,获得了Pareto前沿面曲线和最优解,运用比刚度结构效能对比分析了优化前后的发射筒性能,表明对发射筒的优化设计是有效的。

  7. SLUDGE TREATMENT PROJECT KOP DISPOSITION - THERMAL AND GAS ANALYSIS FOR THE COLD VACUUM DRYING FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON JA; CROWE RD; APTHORPE R; PLYS MG

    2010-03-09

    The purpose of this document is to present conceptual design phase thermal process calculations that support the process design and process safety basis for the cold vacuum drying of K Basin KOP material. This document is intended to demonstrate that the conceptual approach: (1) Represents a workable process design that is suitable for development in preliminary design; and (2) Will support formal safety documentation to be prepared during the definitive design phase to establish an acceptable safety basis. The Sludge Treatment Project (STP) is responsible for the disposition of Knock Out Pot (KOP) sludge within the 105-K West (KW) Basin. KOP sludge consists of size segregated material (primarily canister particulate) from the fuel and scrap cleaning process used in the Spent Nuclear Fuel process at K Basin. The KOP sludge will be pre-treated to remove fines and some of the constituents containing chemically bound water, after which it is referred to as KOP material. The KOP material will then be loaded into a Multi-Canister Overpack (MCO), dried at the Cold Vacuum Drying Facility (CVDF) and stored in the Canister Storage Building (CSB). This process is patterned after the successful drying of 2100 metric tons of spent fuel, and uses the same facilities and much of the same equipment that was used for drying fuel and scrap. Table ES-l present similarities and differences between KOP material and fuel and between MCOs loaded with these materials. The potential content of bound water bearing constituents limits the mass ofKOP material in an MCO load to a fraction of that in an MCO containing fuel and scrap; however, the small particle size of the KOP material causes the surface area to be significantly higher. This relatively large reactive surface area represents an input to the KOP thermal calculations that is significantly different from the calculations for fuel MCOs. The conceptual design provides for a copper insert block that limits the volume available to

  8. EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot

    2001-02-27

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited

  9. 电场对金属串配合物M3(dpa)4Cl2(M=Co,Rh,Ir;dpa=dipyridylamide)结构的影响%Effects of Electric Field on the Structures of Metal String Complexes M3(dpa)4Cl2 (M=Co, Rh, Ir; dpa=dipyridylamide)

    Institute of Scientific and Technical Information of China (English)

    黄燕; 黄晓; 许旋

    2013-01-01

      应用密度泛函理论 PBE0方法研究具有分子导线潜在应用的金属串配合物 M3(dpa)4Cl2(1: M=Co,2:M=Rh,3: M=Ir; dpa=dipyridylamide)在电场作用下的几何和电子结构.结果表明:配合物基态均是二重态.1和2的M36+金属链形成三中心三电子σ键,3中M36+形成三中心四电子σ键且存在弱的δ键.随金属原子周期数增大其M―M键增强、LUMO与HOMO能隙减小、金属原子的反铁磁耦合减弱以至消失且自旋密度向配体的离域增强.在Cl4→Cl5电场作用下,低电势端的M3-Cl5键缩短,高电势端的M2―Cl4键增长, M―M平均键长略为缩短, M―M键增强,有利于分子线的电子传递;分子能量降低,偶极矩线性增大.低电势端Cl5的负电荷向高电势端Cl4转移,且3中金属原子的正电荷由高电势端向低电势端的转移较明显,自旋电子由低电势端向高电势端金属原子移动,但桥联配体dpa-与M和Cl所在的分子轴间没有电荷转移.电场使LUMO与HOMO能隙减小,有利于分子的电子输运.随金属原子周期数增大,电场作用下M―M平均键长变化减小, LUMO、HOMO的能级交错现象减少.%As potential molecular wire species, the geometrical and electronic structures of metal string complexes M3(dpa)4Cl2 (1: M=Co, 2: M=Rh, 3: M=Ir; dpa=dipyridylamide) were investigated theoretical y using density functional theory with the PBE0 functional by considering the interaction of an external electric field along the M36+ linear metal chain. The results show that the ground states of the complexes are al doublets. There is a 3-center-3-electron σ bond delocalized over the M36+ chain for 1 and 2, while there is a 3-center-4-electron σ bond and a weak δ bond among the Ir36+ chain in 3. Moving down the column of Co, Rh, and Ir elements in the periodic table, the complexes with the corresponding metals showed some regular trends, such as stronger M - M bonds, smal er LUMO-HOMO gaps, weaker anti

  10. 新型同心筒自力发射热环境优化设计%Optimization design for thermal environment of a new roadbed concentric canister launcher

    Institute of Scientific and Technical Information of China (English)

    杨风波; 马大为; 任杰; 乐贵高; 蔡德咏

    2015-01-01

    针对新型路基同心筒自力发射热环境评估与优化设计问题,依托弹性变形和域动分层结合的动网格技术,求解了二维轴对称N-S方程,分析了“中段导流”同心筒动态热环境特性,确定了热环境评价指标;通过建立以优化拉丁超立方实验设计和四阶响应面为理论基础的近似数学模型,解决了CFD自动建模困难、直接寻优计算量大的难点;利用多岛遗传和梯度优化算法搭建组合优化策略平台,克服了流场在不同热结构条件下的强非线性问题,并构建了支持近似数学模型的热环境优化构架。对比数值结果表明,倒吸进入内筒的低温气体有力改善了同心筒热环境;建立的近似数学模型精度较高,满足工程需求;优化后,热环境特性发生良性变化,导弹总体热环境得到显著改善。%Based on dynamic mesh technology with spring based smoothing method,and laying based zone moving method, the axisymmetric N-S equations were solved numerically, the dynamic thermal environment characteristics were obtained to deal with the thermal environment evaluating and optimization design problem of the new “middle diversion” concentric canister launcher (CCL),and evaluation index of thermal environment was also determined.The approximate mathematic model was established by op-timal latin hypercube design and fourth-order response surface method to solve the automatic modeling problem of CFD and compen-sate for the shortcoming of large amount of calculation for direct optimization. A combinatorial optimization strategy platform based on Multi-Island Genetic Algorithm and Sequential Quadratic Programming was established to overcome the problem of strong nonlin-ear characteristic of the flow field parameters under different thermal structure conditions,and the optimization design system of ther-mal environment for missile which supports approximate mathematic model was also built

  11. Trade study for the disposition of cesium and strontium capsules

    Energy Technology Data Exchange (ETDEWEB)

    Claghorn, R.D.

    1996-03-01

    This trade study analyzes alternatives for the eventual disposal of cesium and strontium capsules currently stored at the Waste Encapsulation and Storage Facility as by-product. However, for purposes of this study, it is assumed that at some time in the future, the capsules will be declared high-level waste and therefore will require disposal at an offsite geologic repository. The study considered numerous alternatives and selected three for detailed analysis: (1) overpack and storage at high-level waste canister storage building, (2) overpack at the high-level waste vitrification facility followed by storage at a high-level waste canister storage building, and (3) blend capsule contents with other high-level waste feed streams and vitrify at the high-level waste vitrification facility.

  12. L/E coupling numerical simulation of pressure field near launch canister outlet for underwater vehicle vertical launch%水下航行体垂直发射筒口压力场 L/E耦合数值模拟

    Institute of Scientific and Technical Information of China (English)

    张晓乐; 卢丙举; 胡仁海; 杨兴林

    2016-01-01

    The fluid pressure field near canister outlet for underwater-launched vehicle vertical launch was simulated, using a 3D symmetric model based on the coupling of Lagrange structure mesh and Euler fluid mesh. The water horizontal relative motion and the process of vehicle motion in the launch canister were considered in the model. The characteristics of bubble pulsation were achieved through the simulation. The shape of the two primary pressure waves are approximately identical between simulation results and test results. It shows that, simulation model which consider the lateral flow can be more accurate than the model without lateral flow. The research method and its conclusions are good kind of reference to analysis of pressure field near launch canister.%采用 Lagrange 结构网格和 Euler 流场网格(L/E)耦合的数值仿真方法,对潜射航行器出筒后筒口压力场进行三维数值仿真分析。计算模型考虑海水的横向来流和航行体出筒过程对筒口气泡脉动的影响。仿真计算获得了筒口压力场气泡脉动主要特征,试验与计算的2个主要压力波峰曲线形状基本一致。通过对比计算表明,考虑横向流作用可以有效减少筒口压力场计算偏差。本文的计算方法及结论可为发射筒口压力场分析提供有益指导。

  13. Evaluation of Coupled Thermo-Hydro-Mechanical Phenomena in the Near Field for Geological Diaposal of High-Level Radioactive waste

    OpenAIRE

    2000-01-01

    Geological disposal of high-level radioactive waste (HLW) in Japan is based on a multibarrier system composed of engineered and natural barriers. The engineered barriers are composed of vitrified waste confined within a canister, overpack and buffer material. Highly compacted bentonite clay is considered one of the most promising candidate buffer material mainly because of its low hydraunc conductivity and high adsorption capacity of radionuclides. In a repository for HLW, complex thermal, hy...

  14. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  15. Cold Vacuum Drying (CVD) Facility Technical Safety Requirements

    Energy Technology Data Exchange (ETDEWEB)

    KRAHN, D.E.

    2000-08-08

    The Technical Safety Requirements (TSRs) for the Cold Vacuum Drying Facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt of multi-canister overpacks (MCOs) containing spent nuclear fuel. removal of free water from the MCOs using the cold vacuum drying process, and inerting and testing of the MCOs before transport to the Canister Storage Building. Controls required for public safety, significant defense in depth, significant worker safety, and for maintaining radiological and toxicological consequences below risk evaluation guidelines are included.

  16. Preliminary conceptual designs for advanced packages for the geologic disposal of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Westerman, R.E.

    1979-04-01

    The present study assumes that the spent fuel will be disposed of in mined repositories in continental geologic formations, and that the post-emplacement control of the radioactive species will be accomplished independently by both the natural barrier, i.e., the geosphere, and the engineered barrier system, i.e., the package components consisting of the stabilizer, the canister, and the overpack; and the barrier components external to the package consisting of the hole sleeve and the backfill medium. The present document provides an overview of the nature of the spent fuel waste; the general approach to waste containment, using the defense-in-depth philosophy; material options, both metallic and nonmetallic, for the components of the engineered barrier system; a set of strawman criteria to guide the development of package/engineered barrier systems; and four preliminary concepts representing differing approaches to the solution of the containment problem. These concepts use: a corrosion-resistant meta canister in a special backfill (2 barriers); a mild steel canister in a corrosion-resistant metallic or nonmetallic hole sleeve, surrounded by a special backfill (2 barriers); a corrosion-resistant canister and a corrosion-resistant overpack (or hole sleeve) in a special backfill (3 barriers); and a mild steel canister in a massive corrosion-resistant bore sleeve surrounded by a polymer layer and a special backfill (3 barriers). The lack of definitive performance requirements makes it impossible to evaluate these concepts on a functional basis at the present time.

  17. Sealed Planetary Return Canister (SPRC) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Sample return missions have primary importance in future planetary missions. A basic requirement is that samples be returned in pristine, uncontaminated condition,...

  18. Draft Geologic Disposal Requirements Basis for STAD Specification

    Energy Technology Data Exchange (ETDEWEB)

    Ilgen, Anastasia G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-25

    This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussed here.

  19. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-09-30

    The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration.

  20. Engineered design features in the HI-STAR/HI-STORM systems to maximize ALARA, safety, and community acceptance

    Energy Technology Data Exchange (ETDEWEB)

    Blessing, Christian [Holtec International, New Jersey (United States)

    2003-07-01

    Heltec International is a U.S. corporation headquartered in New Jersey, dedicated to providing capital goods and technical services to the power industry. Over 75 percent of the company's product output is destined for nuclear power plants. Holter counts among its active clients a majority of the nuclear plants in the United States, as well as Korea, Taiwan, Mexico, and Brazil. The company also has a growing market presence in Japan and the European Union. Leading U.S. nuclear plant owners, such as Entergy, Exelon, FPL, Southern Nuclear, PG and E and TVA have a long-term and continuous business relationship with Holtec International. This article deals with Holtec dry storage system description, the multi-purpose canister, hi-star 100 overpack, hi-storm 100 overpack and unique advantages of holtec's dry storage technology.

  1. Cold Vacuum Drying Facility Crane and Hoist System Design Description (SYS 14)

    Energy Technology Data Exchange (ETDEWEB)

    TRAN, Y.S.

    2000-06-07

    This system design description (SDD) is for the Cold Vacuum Drying (CVD) Facility overhead crane and hoist system. The overhead crane and hoist system is a general service system. It is located in the process bays of the CVD Facility, supports the processes required to drain the water and dry the spent nuclear fuel (SNF) contained in the multi-canister overpacks (MCOs) after they have been removed from the K-Basins. The location of the system in the process bay is shown.

  2. Retrievable storage concept designs. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E.

    1979-03-01

    Three tasks related to the reference design of retrievable storage canisters for radioactive waste have been completed. The three tasks consist of the reference design itself, the definition of failure modes most appropriate for structural integrity determinations for the reference canister, and the development of a failure methodology for the structural integrity of the containers. The reference design is a sealed storage canister concept based upon the waste isolation pilot plant (WIPP) design, with slight modifications. The modifications consist of an alternate lifting yoke arrangement for the top head and a revised bottom head design for absorption of impact energy. Welded closures provide the seal at each end. Overpacking is considered as a possibility, but is not included in the preliminary reference design. The four failure modes that are deemed the most appropriate for the design of the reference canister are: (i) a loss of functional capability; (ii) ductile rupture of the canister; (iii) buckling of the structural members; and (iv) stress corrosion cracking. Failure scenarios are provided for each of the relevant failure modes. In addition, a failure methodology based upon the distribution of demand and the distribution of capacity for the structural members, with respect to each failure mode, is proffered.

  3. Biosynthesis of the Cyclotide MCoTI-II using an Engineered Intein

    Energy Technology Data Exchange (ETDEWEB)

    Cantor, J; Camarero, J A

    2006-08-15

    Cyclotides are an emerging family of naturally occurring circular mini-proteins ({approx}30-40 amino acids) characterized by six conserved Cys residues (forming 3 disulfide bridges) that create a topologically unique structure designated as a cyclic cysteine knot (CCK). The cysteine knot motif, which is embedded within the macrocylic backbone, is described as two disulfide bridges that form a ring that is penetrated by the third disulfide bridge. The cyclic backbone and CCK motif together confer cyclotides with a remarkable stability and resistance to proteolytic, chemical, and thermal degradation. Further, cyclotides are functionally diverse and display a wide range of functions including uterotonic activity, trypsin inhibition, cytotoxicity, neurotensin binding, anti-HIV, antimicrobial, and insecticidal activity. Together, these characteristics make cyclotides attractive candidates for both drug design and agricultural applications, both in their native forms and as molecular scaffolds for the incorporation of novel bioactivities. [1] The ability to manipulate production of cyclotides within biological systems is critical for mutagenesis studies, production of grafted products, and the mass production of cyclotides with novel activities. My adviser's hope is to achieve this capability by employing recombinant DNA expression techniques to generate large combinatorial libraries of cyclotides. The advantage in creating a biosynthetic library (containing {approx}10{sup 6}-10{sup 10} members/library vs. chemically based libraries with typical values ranging from {approx}10{sup 3}-10{sup 5} members/library) is that it can be lead to the in vivo application of biological screening and selection methodologies based on a specific clone's ability to affect certain cellular processes.

  4. 42 CFR 84.1154 - Canister and cartridge requirements.

    Science.gov (United States)

    2010-10-01

    ..., and Mist; Pesticide; Paint Spray; Powered Air-Purifying High Efficiency Respirators and Combination... are used in parallel, their resistance to airflow shall be essentially equal. (b) The color...

  5. Phase 2 fire hazard analysis for the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Sadanaga, C.T., Westinghouse Hanford

    1996-07-01

    The fire hazard analysis assesses the risk from fire in a facility to ascertain whether the fire protection policies are met. This document provides a preliminary FHA for the CSB facility. Open items have been noted in the document. A final FHA will be required at the completion of definitive design, prior to operation of the facility.

  6. Modellization of Metal Hydride Canister for Hydrogen Storage

    Directory of Open Access Journals (Sweden)

    Rocio Maceiras

    2015-06-01

    Full Text Available Hydrogen shows very interesting features for its use on-board applications as fuel cell vehicles. This paper presents the modelling of a tank with a metal hydride alloy for on-board applications, which provides good performance under ambient conditions. The metal hydride contained in the tank is Ti0.98Zr0.02V0.43Fe0.09Cr0.05Mn1.5. A two-dimensional model has been performed for the refuelling process (absorption and the discharge process (desorption. For that, individual models of mass balance, energy balance, reaction kinetics and behaviour of hydrogen gas has been modelled. The model has been developed under Matlab / Simulink© environment. Finally, individual models have been integrated into a global model, and simulated under ambient conditions. With the aim to analyse the temperature influence on the state of charge and filling and emptying time, other simulations were performed at different temperatures. The obtained results allow to conclude that this alloy offers a good behaviour with the discharge process under normal ambient conditions. Keywords: Hydrogen storage; metal hydrides; fuel cell; simulation; board applications

  7. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 22. Nuclear considerations for repository design

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

    This volume, Y/OWI/TM-36/22, ''Nuclear Considerations for Repository Design,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. Included in this volume are baseline design considerations such as characteristics of canisters, drums, casks, overpacks, and shipping containers; maximum allowable and actual decay-heat levels; and canister radiation levels. Other topics include safeguard and protection considerations; occupational radiation exposure including ALARA programs; shielding of canisters, transporters and forklift trucks; monitoring considerations; mine water treatment; canister integrity; and criticality calculations.

  8. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  9. Analysis of Dust Samples Collected from an Unused Spent Nuclear Fuel Interim Storage Container at Hope Creek, Delaware.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    In July, 2014, the Electric Power Research Institute and industry partners sampled dust on the surface of an unused canister that had been stored in an overpack at the Hope Creek Nuclear Generating Station for approximately one year. The foreign material exclusion (FME) cover that had been on the top of the canister during storage, and a second recently - removed FME cover, were also sampled. This report summarizes the results of analyses of dust samples collected from the unused Hope Creek canister and the FME covers. Both wet and dry samples of the dust/salts were collected, using SaltSmart(TM) sensors and Scotch - Brite(TM) abrasive pads, respectively. The SaltSmart(TM) samples were leached and the leachate analyzed chemically to determine the composition and surface load per unit area of soluble salts present on the canister surface. The dry pad samples were analyzed by X-ray fluorescence and by scanning electron microscopy to determine dust texture and mineralogy; and by leaching and chemical analysis to deter mine soluble salt compositions. The analyses showed that the dominant particles on the canister surface were stainless steel particles, generated during manufacturing of the canister. Sparse environmentally - derived silicates and aluminosilicates were also present. Salt phases were sparse, and consisted of mostly of sulfates with rare nitrates and chlorides. On the FME covers, the dusts were mostly silicates/aluminosilicates; the soluble salts were consistent with those on the canister surface, and were dominantly sulfates. It should be noted that the FME covers were w ashed by rain prior to sampling, which had an unknown effect of the measured salt loads and compositions. Sulfate salts dominated the assemblages on the canister and FME surfaces, and in cluded Ca - SO4 , but also Na - SO4 , K - SO4 , and Na - Al - SO4 . It is likely that these salts were formed by particle - gas conversion reactions, either

  10. A Facile Synthesis of MPd (M=Co, Cu) Nanoparticles and Their Catalysis for Formic Acid Oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Mazumder, Vismadeb [Brown University; Chi, Miaofang [ORNL; Mankin, Max [Brown University; Liu, Yi [Brown University; Metin, Onder [Ataturk University; Sun, Daohua [Xiamen University, China; More, Karren Leslie [ORNL; Sun, Shouheng [Brown University

    2012-01-01

    Monodisperse CoPd nanoparticles (NPs) were synthesized and studied for catalytic formic acid (HCOOH) oxidation (FAO). The NPs were prepared by coreduction of Co(acac)2 (acac = acetylacetonate) and PdBr2 at 260 C in oleylamine and trioctylphosphine, and their sizes (5-12 nm) and compositions (Co10Pd90 to Co60Pd40) were controlled by heating ramp rate, metal salt concentration, or metal molar ratios. The 8 nm CoPd NPs were activated for HCOOH oxidation by a simple ethanol wash. In 0.1 M HClO4 and 2 M HCOOH solution, their catalytic activities followed the trend of Co50Pd50 > Co60Pd40 > Co10Pd90 > Pd. The Co50Pd50 NPs had an oxidation peak at 0.4 V with a peak current density of 774 A/gPd. As a comparison, commercial Pd catalysts showed an oxidation peak at 0.75 V with peak current density of only 254 A/gPd. The synthesis procedure could also be extended to prepare CuPd NPs when Co(acac)2 was replaced by Cu(ac)2 (ac = acetate) in an otherwise identical condition. The CuPd NPs were less active catalysts than CoPd or even Pd for FAO in HClO4 solution. The synthesis provides a general approach to Pd-based bimetallic NPs and will enable further investigation of Pd-based alloy NPs for electro-oxidation and other catalytic reactions.

  11. Electrochemical properties of LaMO3 (M=Co or Fe) as the negative electrode in a hydrogen battery

    Science.gov (United States)

    Lim, D.-K.; Im, H.-N.; Kim, J.; Song, S.-J.

    2013-01-01

    Undoped orthorthombic LaFeO3 and monoclinic LaCoO3 oxides were selected as an anode material for Ni-H battery due to their high electron conductivity by multivalent transition status of B-site cation. Both groups of oxides were prepared by a conventional solid-state reaction method, and their electrochemical charge/discharge properties were investigated. The electrochemical kinetic properties, exchange current density, and proton diffusivity were also extracted using linear polarization measurement and the potential-step method. X-ray photoelectron spectroscopy (XPS) analysis was used to measure the oxidation state of the transition metal in the specimens. A non-linear least-square fitting deconvoluted the peaks, suggesting that the valence state of Fe and Co in the sample was mainly +3. The hydrogen diffusion rate was also estimated using the potential-step method, giving 5.42×10-16 and 5.72×10-16 cm2 s-1 for LaCoO3 and LaFeO3, respectively which are an order of magnitude larger than that of Sr doped LaFeO3 oxide electrodes.

  12. Theoretical study of the multiferroic properties in M-doped (M=Co, Cr, Mg) ZnO thin films

    Energy Technology Data Exchange (ETDEWEB)

    Bahoosh, S.G. [Max Planck Institute of Microstructure Physics, Weinberg 2, 06120 Halle (Germany); Apostolov, A.T. [University of Architecture, Civil Engineering and Geodesy, Faculty of Hydrotechnics, Department of Physics, 1, Hristo Smirnenski Blvd., 1046 Sofia (Bulgaria); Apostolova, I.N. [University of Forestry, Faculty of Forest Industry, 10, Kl. Ohridsky Blvd., 1756 Sofia (Bulgaria); Trimper, S. [Institute of Physics, Martin-Luther-University, D-06099 Halle (Germany); Wesselinowa, Julia M. [University of Sofia, Department of Physics, Blvd. J. Bouchier 5, 1164 Sofia (Bulgaria)

    2015-01-01

    The origin of multiferroism is still an open problem in ZnO. We propose a microscopic model to clarify the occurrence of multiferroism in this material. Using Green's function technique we study the influence of ion doping and size effects on the magnetization and polarization of ZnO thin films. The calculations for magnetic Co- and Cr-ions are based on the s–d model, the transverse Ising model in terms of pseudo-spins and a biquadratic magnetoelectric coupling, whereas in case of nonmagnetic Mg-ions the model takes into account the Coulomb interaction and an indirect coupling between the pseudo-spins via the conduction electrons. We show that the magnetization M exhibits a maximum for a fixed concentration of the doping ions. Furthermore M increases with decreasing film thickness N. The polarization increases with increasing concentration of the dopant and decreasing N. The results are in good agreement with the experimental data. - Highlights: • The paper analyzes the multiferroic properties of doped ZnO thin films by a microscopic model. • The magnetization exhibits a maximum at a fixed doping concentration. • The polarization increases with growing dopant concentration. • The ferroelectric transition temperature is enhanced for increasing dopant concentration.

  13. Activity measurement of 60Fe through the decay of 60mCo and confirmation of its half-life

    CERN Document Server

    Ostdiek, Karen; Bauder, William; Bowers, Matthew; Clark, Adam; Collon, Philippe; Dressler, Rugard; Greene, John; Kutschera, Walter; Lu, Wenting; Nelson, Austin; Paul, Michael; Robertson, Daniel; Schumann, Dorothea; Skulski, Michael

    2016-01-01

    The half-life of the neutron-rich nuclide, $^{60}\\text{Fe}$ has been in dispute in recent years. A measurement in 2009 published a value of $(2.62 \\pm 0.04)\\times10^{6}$ years, almost twice that of the previously accepted value from 1984 of $(1.49 \\pm 0.27)\\times10^{6}$ years. This longer half-life was confirmed in 2015 by a new measurement, resulting in a value of $(2.50 \\pm 0.12)\\times10^{6}$ years. All three half-life measurements used the grow-in of the $\\gamma$-ray lines in $^{60}\\text{Ni}$ from the decay of the ground state of $^{60\\text{g}}\\text{Co}$ (t$_{1/2}$=5.27 years) to determine the activity of a sample with a known number of $^{60}\\text{Fe}$ atoms. In contrast, the work presented here measured the $^{60}\\text{Fe}$ activity directly with the 58.6 keV $\\gamma$-ray line from the short-lived isomeric state of $^{60\\text{m}}\\text{Co}$ (t$_{1/2}$=10.5 minutes), thus being independent of any possible contamination from long-lived $^{60\\text{g}}\\text{Co}$. A fraction of the material from the 2015 exper...

  14. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R.; Enos, David George

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  15. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David George [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  16. Conceptual Design Report Cask Loadout Sys and Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform at 105 K West

    Energy Technology Data Exchange (ETDEWEB)

    LANGEVIN, A.S.

    1999-07-12

    This conceptual design report documents the redesign of the IPSS and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5/A.6, Canister Transfer Facility Modifications. Project A.5/A.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The function of the IPSS is to suspend, guide, and position the immersion pail. The immersion pail protects the cask-MCO from contamination by basin water and acts as a lifting device for the cask-MCO. The OIP provides operator access to the south loadout pit. Previous analyses studied the effects of a cask-MCO drop on the south loadout pit concrete structure and on the IPSS. The most recent analysis considered the resulting loads at the pit slab/wall joint (Kanjilal, 1999). This area had not been modeled previously, and the analysis results indicate that the demand capacity exceeds the allowable at the slab/wall joint. The energy induced on the south loadout pit must be limited such that the safety class function of the basin is maintained. The solution presented in this CDR redesigns the IPSS and the OIP to include impact-absorbing features that will reduce the induced energy. The impact absorbing features of the new design include: Impact-absorbing material at the IPSS base and at the upper portion of the IPSS legs. A sleeve which provides a hydraulic means of absorbing energy. Designing the OIP to act as an impact absorber. The existing IPSS structure in 105 KW will be removed. This conceptual design considers only loads resulting from drops directly over the IPSS and south loadout pit area. Drops in other areas of the basin are not considered, and will be covered as part of a future revision to this CDR.

  17. Spent nuclear fuel project detonation phenomena of hydrogen/oxygen in spent fuel containers

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, T.D.

    1996-09-30

    Movement of Spent N Reactor fuels from the Hanford K Basins near the Columbia River to Dry interim storage facility on the Hanford plateau will require repackaging the fuel in the basins into multi-canister overpacks (MCOs), drying of the fuel, transporting the contained fuel, hot conditioning, and finally interim storage. Each of these functions will be accomplished while the fuel is contained in the MCOs by several mechanisms. The principal source of hydrogenand oxygen within the MCOs is residual water from the vacuum drying and hot conditioning operations. This document assesses the detonation phenomena of hydrogen and oxygen in the spent fuel containers. Several process scenarios have been identified that could generate detonation pressures that exceed the nominal 10 atmosphere design limit ofthe MCOS. Only 42 grams of radiolized water are required to establish this condition.

  18. Materials Characterization Center state-of-the-art report on corrosion data pertaining to metallic barriers for nuclear-waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Merz, M.D.

    1982-10-01

    A compilation of published corrosion data on metals that have been suggested as canisters and overpack materials is presented. The data were categorized according to the solutions used in testing and divided into two parts: high-ionic strength solutions (such as seawater and brine) and low-ionic-strength waters (such as basalt and tuff waters). This distinction was made primarily because of the general difference in aggressiveness of these solutions with respect to general corrosion. A considerable amount of data indicated that titanium alloys have acceptably low uniform corrosion rates in anticipated repository sites; the other possible corrosion failure modes for titanium alloys, such as stress corrosion cracking and delayed failure due to hydrogen, have not been sufficiently studied to make any similar conclusions about lifetime with respect to these particular degradation processes. Other data suggested that iron-base alloys are sufficiently resistant to corrosion in basalt and tuff waters, although the effects of radiation and radiation combined with elevated temperature have not been reported in enough detail to conclusively qualify iron-base alloys for any particular barrier thickness in regard to uniform corrosion rate. The effect of overpack size on corrosion rate has been given little attention. A review of long-term underground data indicated that temperature and accessibility to oxygen were too different for deep geologic repositories to make the underground corrosion data directly applicable. However, the characteristics of corrosion attack, statistical treatment of data, and kinetics of corrosion showed that corrosion proceeds in a systematic and predictable way.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Assessment of radionuclide release scenarios for the repository system 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Assessment of Radionuclide Release Scenarios sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of presenting an assessment of the repository system scenarios leading to radionuclide releases that have been identified in Formulation of Radionuclide Release Scenarios. A base scenario, variant scenarios and disturbance scenarios are considered. For each scenario, a range of calculation cases, also identified in Formulation of Radionuclide Release Scenarios, has been analysed, complemented by Monte Carlo simulations, a probabilistic sensitivity analysis and other supporting calculations. The calculation cases and analyses take into account major uncertainties in the initial state of the barriers and possible paths for the evolution of the repository system identified in Performance Assessment. Quality control and assurance measures have been adopted to ensure transparency and traceability of the calculations performed and hence to promote confidence in the analysis of the calculation cases. The calculation cases each consider a single, failed canister, where three possible modes of failure are addressed: (1) the presence of an initial penetrating defect in the copper overpack of the canister, (2) corrosion of the copper overpack, which occurs most rapidly in scenarios in which buffer density is reduced, e.g. by erosion, (3) shear movement on a fracture intersecting a deposition hole. The likelihood and consequences of multiple canister failure occurring during the assessment time frame are also considered. In particular, the analyses consider: The likelihood and consequences of there being multiple canisters with initial penetrating defects; The consequences if canister failure due to corrosion following buffer erosion were to occur; and The low annual probability of there being an earthquake large enough to give rise to canister failure due to rock shear movements and the potential consequences of such an earthquake

  1. Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Kalinina, Elena Arkadievna; Hardin, Ernest

    2015-11-01

    This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assume that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.

  2. Clay Generic Disposal System Model - Sensitivity Analysis for 32 PWR Assembly Canisters (+2 associated model files).

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Edgar [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-10-01

    The Used Fuel Disposition Campaign (UFDC), as part of the DOE Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technology program (FCT) is investigating the disposal of high level radioactive waste (HLW) and spent nuclear fuela (SNF) in a variety of geologic media. The feasibility of disposing SNF and HLW in clay media has been investigated and has been shown to be promising [Ref. 1]. In addition the disposal of these wastes in clay media is being investigated in Belgium, France, and Switzerland. Thus, Argillaceous media is one of the environments being considered by UFDC. As identified by researchers at Sandia National Laboratory, potentially suitable formations that may exist in the U.S. include mudstone, clay, shale, and argillite formations [Ref. 1]. These formations encompass a broad range of material properties. In this report, reference to clay media is intended to cover the full range of material properties. This report presents the status of the development of a simulation model for evaluating the performance of generic clay media. The clay Generic Disposal System Model (GDSM) repository performance simulation tool has been developed with the flexibility to evaluate not only different properties, but different waste streams/forms and different repository designs and engineered barrier configurations/ materials that could be used to dispose of these wastes.

  3. The Use of One-Sample Prediction Intervals for Estimating CO2 Scrubber Canister Durations

    Science.gov (United States)

    2012-10-01

    pp. 299–301. 5. R. L. Scheaffer and J. T. McClave, Probability and Statistics, 3rd ed. (Boston, MA: PWS-Kent, 1990), pp. 291–292. 6. G. G...1990), p. 82. 2. Ibid., p. 206. 14 B-1 Appendix B: Textbook Examples 1. Scheaffer and McClave (1990), p. 292. Ten independent observations...next observation will lie between 16.076 and 16.124. 2. Scheaffer and McClave (1990), p. 293. A particular subcompact automobile has been tested

  4. Canister storage building compliance assessment DOE Order 6430.1A, General Design Criteria

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.

    1999-08-12

    This document presents the Project's position on compliance with DOE Order 6430.1A ''General Design Criteria.'' No non-compliances are shown. The compliance statements have been reviewed and approved by DOE. Open items are scheduled to be closed prior to project completion.

  5. 40 CFR 86.153-98 - Vehicle and canister preconditioning; refueling test.

    Science.gov (United States)

    2010-07-01

    ... part). Fifteen seconds after the engine starts, place the transmission in gear. Twenty seconds after... controlled to 50±25 grains of water vapor per pound of dry air) maintained at a nominal flow rate of 0.8 cfm... vapor per pound of dry air) maintained at a nominal flow rate of 0.8 cfm. In this case, the...

  6. Canister storage building compliance assessment SNF project NRC equivalency criteria - HNF-SD-SNF-DB-003

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.

    1999-08-11

    This document presents the Project's position on compliance with the SNF Project NRC Equivalency Criteria--HNF-SD-SNF-DE-003, Spent Nuclear Fuel Project Path Forward Additional NRC Requirements. No non-compliances are shown The compliance statements have been reviewed and approved by DOE. Open items are scheduled to be closed prior to project completion.

  7. Stress corrosion cracking in canistered waste package containers: Welds and base metals

    Energy Technology Data Exchange (ETDEWEB)

    Huang, J.S.

    1998-03-01

    The current design of waste package containers include outer barrier using corrosion allowable material (CAM) such as A516 carbon steel and inner barrier of corrosion resistant material (CRM) such as alloy 625 and C22. There is concern whether stress corrosion cracking would occur at welds or base metals. The current memo documents the results of our analysis on this topic.

  8. A REVIEW OF THE US EPA'S SINGLE BREATH CANISTER (SBC) METHOD FOR EXHALED VOLATILE ORGANIC BIOMARKERS

    Science.gov (United States)

    Exhaled alveolar breath can provide a great deal of information about an individual?s health and previous exposure to potentially harmful xenobiotic materials. Because breath can be obtained noninvasively and its constituents directly reflect concentrations in the blood, its us...

  9. Developing a structural health monitoring system for nuclear dry cask storage canister

    Science.gov (United States)

    Sun, Xiaoyi; Lin, Bin; Bao, Jingjing; Giurgiutiu, Victor; Knight, Travis; Lam, Poh-Sang; Yu, Lingyu

    2015-03-01

    Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. In total, there are over 1482 dry cask storage system (DCSS) in use at US plants, storing 57,807 fuel assemblies. Nondestructive material condition monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health", and more importantly, to guarantee the safe operation of radioactive waste storage systems (RWSS) during their extended usage period. A state-of-the-art nuclear structural health monitoring (N-SHM) system based on in-situ sensing technologies that monitor material degradation and aging for nuclear spent fuel DCSS and similar structures is being developed. The N-SHM technology uses permanently installed low-profile piezoelectric wafer sensors to perform long-term health monitoring by strategically using a combined impedance (EMIS), acoustic emission (AE), and guided ultrasonic wave (GUW) approach, called "multimode sensing", which is conducted by the same network of installed sensors activated in a variety of ways. The system will detect AE events resulting from crack (case for study in this project) and evaluate the damage evolution; when significant AE is detected, the sensor network will switch to the GUW mode to perform damage localization, and quantification as well as probe "hot spots" that are prone to damage for material degradation evaluation using EMIS approach. The N-SHM is expected to eventually provide a systematic methodology for assessing and monitoring nuclear waste storage systems without incurring human radiation exposure.

  10. Role of Bismuth Oxide in Bi-MCo2O4(M=Co,Ni,Cu,Zn) Catalysts for Wet Air Oxidation of Acetic Acid

    Institute of Scientific and Technical Information of China (English)

    JIANG Peng-bo; CHENG Tie-xin; ZHUANG Hong; CUI Xiang-hao; BI Ying-li; ZHEN Kai-ji

    2004-01-01

    Two series of cobalt(Ⅲ)-containing spinel catalysts were prepared by the decomposition of the corresponding nitrates. The catalysts doped with bismuth oxide exhibit a higher activity in the wet air oxidation of acetic acid than those without dopant bismuth oxide. The catalysts were investigated by XRD, TEM, ESR, UV-DRS and XPS, and the interaction between Co and Bi was studied as well. It has been found that nano-sized bismuth oxide is paved on the surface of cobalt spinel crystal and the structures of cobalt(Ⅲ)-containing spinel are still maintained. The shift of the binding energy of Bi4f7/2 is related to the catalytic activity of these catalysts doped with bismuth oxide.

  11. 宫颈癌调强放疗 MCO 与 DMPO 算法的计划剂量学比较

    Institute of Scientific and Technical Information of China (English)

    张国前; 张书旭; 余辉; 谭剑明; 王锐濠; 雷怀宇; 蒋绍惠; 周祥

    2016-01-01

    目的:比较宫颈癌调强放疗时不同优化算法( MCO与DMPO)所得计划的剂量学差异。方法将收治的肿瘤患者中随机抽样选取10例接受根治术后的Ⅰ~Ⅱ期宫颈癌病例,依次设计4组计划:用MCO算法分别设计M-IMRT和M-VMAT;用DMPO算法分别设计多野调强计划( D-IMRT)和容积调强计划( D-VMAT);比较两种不同优化算法所生成IMRT或VMAT计划中靶区、危及器官的剂量分布及机器跳数的差异。结果(1) M-IMRT与D-IMRT相比,其靶区D95、靶区Dmax、靶区Dmin分别增加1.0%(P=0.02)、2.0%(P=0.001)、1.6%(P=0.04),适型指数降低4.5%(P=0.02)。(2)M-VMAT总跳数比D-VMAT增加13.8%(t =7.26,P=0.01)。(3)除M-VMAT计划右侧股骨头V40增加以外其余MCO计划中直肠、膀胱及双侧股骨头剂量相比DMPO计划均有不同程度的降低。结论在满足靶区剂量的基础上MCO算法能够降低危及器官受照剂量,但结果还需进一步验证。

  12. Synthesis, characterization and magnetic properties of MFe{sub 2}O{sub 4} (M=Co, Mg, Mn, Ni) nanoparticles using ricin oil as capping agent

    Energy Technology Data Exchange (ETDEWEB)

    Gherca, Daniel [Faculty of Chemistry, Alexandru Ioan Cuza University, Blvd. Carol I nr 11, Iasi 700506 (Romania); Pui, Aurel, E-mail: aurel@uaic.ro [Faculty of Chemistry, Alexandru Ioan Cuza University, Blvd. Carol I nr 11, Iasi 700506 (Romania); Cornei, Nicoleta [Faculty of Chemistry, Alexandru Ioan Cuza University, Blvd. Carol I nr 11, Iasi 700506 (Romania); Cojocariu, Alina; Nica, Valentin; Caltun, Ovidiu [Faculty of Physics and Carpath Center, Alexandru Ioan Cuza University, Blvd. Carol I nr 11, Iasi 700506 (Romania)

    2012-11-15

    We focused on obtaining MFe{sub 2}O{sub 4} nanoparticles using ricin oil solution as surfactant and on their structural characterization and magnetic properties. The annealed samples at 500 Degree-Sign C in air for 6 h were analyzed for the crystal phase identification by powder X-ray diffraction using CuK{alpha} radiation. The particle size, the chemical composition and the morphology of the calcinated powders were characterized by scanning electron microscopy. All sintered samples contain only one phase, which has a cubic structure with crystallite sizes of 12-21 nm. From the infrared spectra of all samples were observed two strong bands around 600 and 400 cm{sup -1}, which correspond to the intrinsic lattice vibrations of octahedral and tetrahedral sites of the spinel structure, respectively, and characteristic vibration for capping agent. The magnetic properties of fine powders were investigated at room temperature by using a vibrating sample magnetometer. The room temperature M-H hysteresis loops show ferromagnetic behavior of the calcined samples, with specific saturation magnetization (M{sub s}) values ranging between 11 and 53 emu/g. - Highlights: Black-Right-Pointing-Pointer MFe{sub 2}O{sub 4} nanoparticles obtained using ricin oil as surfactant. Black-Right-Pointing-Pointer The structures were confirmed by SEM micrographs, FTIR spectra and XRD spectroscopy. Black-Right-Pointing-Pointer The samples contain only one phase with crystallite sizes of 12-21 nm.

  13. Intermolecular hydrogen bonding between neutral transition metal hydrides (eta(5)-C5H5)M(CO)3H (M = Mo, W) and bases.

    Science.gov (United States)

    Belkova, Natalia V; Gutsul, Evgenii I; Filippov, Oleg A; Levina, Vladislava A; Valyaev, Dmitriy A; Epstein, Lina M; Lledos, Agusti; Shubina, Elena S

    2006-03-22

    The interaction of CpM(CO)3H (M = Mo, W) hydrides as proton donors with different bases (B = pyridine, (n-Oc)3PO, ((CH3)2N)3PO, H3BNEt3) was studied by variable temperature IR spectroscopy and theoretically by DFT/B3LYP calculations. The data obtained show for the first time the formation of intermolecular hydrogen bonds between the neutral transition metal hydrides and bases in solutions of low polarity. These M-H...B hydrogen bonds are shown to precede the hydrides' deprotonation.

  14. 42 CFR 438.56 - Disenrollment: Requirements and limitations.

    Science.gov (United States)

    2010-10-01

    ... enrollees); and (3) Specify the methods by which the MCO, PIHP, PAHP, or PCCM assures the agency that it... enrollee seek redress through the MCO, PIHP, PAHP, or PCCM's grievance system before making a determination... MCO, PIHP, PAHP, or PCCM files the request. (2) If the MCO, PIHP, PAHP, or PCCM or the State...

  15. Biocatalytic potential of laccase-like multicopper oxidases from Aspergillus niger

    Directory of Open Access Journals (Sweden)

    Tamayo-Ramos Juan Antonio

    2012-12-01

    Full Text Available Abstract Background Laccase-like multicopper oxidases have been reported in several Aspergillus species but they remain uncharacterized. The biocatalytic potential of the Aspergillus niger fungal pigment multicopper oxidases McoA and McoB and ascomycete laccase McoG was investigated. Results The laccase-like multicopper oxidases McoA, McoB and McoG from the commonly used cell factory Aspergillus niger were homologously expressed, purified and analyzed for their biocatalytic potential. All three recombinant enzymes were monomers with apparent molecular masses ranging from 80 to 110 kDa. McoA and McoG resulted to be blue, whereas McoB was yellow. The newly obtained oxidases displayed strongly different activities towards aromatic compounds and synthetic dyes. McoB exhibited high catalytic efficiency with N,N-dimethyl-p-phenylenediamine (DMPPDA and 2,2-azino-di(3-ethylbenzthiazoline sulfonic acid (ABTS, and appeared to be a promising biocatalyst. Besides oxidizing a variety of phenolic compounds, McoB catalyzed successfully the decolorization and detoxification of the widely used textile dye malachite green. Conclusions The A. niger McoA, McoB, and McoG enzymes showed clearly different catalytic properties. Yellow McoB showed broad substrate specificity, catalyzing the oxidation of several phenolic compounds commonly present in different industrial effluents. It also harbored high decolorization and detoxification activity with the synthetic dye malachite green, showing to have an interesting potential as a new industrial biocatalyst.

  16. Expansion due to the anaerobic corrosion of iron

    Energy Technology Data Exchange (ETDEWEB)

    Smart, N.R.; Rance, A.P.; Fennell, P.A.H. [Serco Assurance, Culham Science Centre (United Kingdom)

    2006-12-15

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in vertical storage holes drilled in a series of caverns excavated from the granite bedrock at a depth of about 500 m and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container, designed to provide mechanical strength and to keep individual fuel bundles at a safe distance from one another, thereby minimising the risk of criticality. The container is fitted inside an inherently corrosion resistant copper overpack that is designed to provide containment over the long timescales required. As part of the safety case for the repository, one of the scenarios being addressed by SKB involves the early mechanical failure of the outer copper overpack, allowing water to enter the outer container and corrode the inner one. One consequence of this failure would be the long-term build up of corrosion product, which could induce stresses in the spent fuel canister. A programme of experimental work was undertaken to investigate the effect of corrosion product formation on the generation of stresses in the outer copper container. This report describes the construction of an apparatus to directly measure the expansion caused by the anaerobic corrosion of ferrous material in a simulated repository environment whilst under representative compressive loads. This apparatus, known as the 'stress cell' consisted of a stack of interleaved carbon steel and copper discs that was subjected to a compressive load simulating the loads expected in a repository and immersed in simulated anoxic groundwater at 69 deg C. The stack was mounted in a rigid frame and a system of levers was used to amplify any expansion caused by corrosion; the expansion of the stack was measured using sensitive displacement transducers

  17. Review of high-level waste form properties. [146 bibliographies

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.M.

    1980-12-01

    This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.

  18. Basalt Waste Isolation Project. Quarterly report, January 1-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Deju, R.A.

    1980-04-01

    This report addresses the technical progress for the Basalt Waste Isolation Project for the second quarter of fiscal year 1980. Seismic design values were developed for preliminary repository design purposes; 0.25 g horizontal and 0.125 g vertical maximum accelerations for surface, zero-period conditions. Preliminary seismic data indicate broad, smooth areas exist in the bedrock surface in the western portion of the Cold Creek syncline and a gently undulating bedrock surface in the eastern portion. Test results indicate hydraulic property values fall within the range previously reported for sedimentary and interflow zones in basalt formations at the Hanford Site. Preliminary results of available hydrochemical data obtained from several borehole sites indicate that little, if any, vertical mixing of groundwaters is taking place across this stratigraphic boundary. Multiple barrier studies indicate that the primary candidate canister/overpack alloys are TiCode-12, Inconel 625, Incoloy 825, and Zircaloy 2. Low-carbon steel and cast iron are among the list of secondary candidate canister alloys. Laboratory tests of borehole plug designs have shown that it is feasible to design a composite plug system that will satisfactorily seal a nuclear waste repository in Columbia River basalt. The National Lead Industries, Inc., NLI-1/2 Universal Spent Fuel Shipping Cask was selected for use in Phase II operations. Creep test results of samples of Umtanum basalt from borehole DC-6 were plotted and show the day-to-day variation in deformation versus time. The concept selection phase of repository conceptual design was completed in March 1980. A test plan for the Exploratory Shaft Test Facility was developed and is scheduled for submittal to the US Department of Energy in May 1980.

  19. Status report. Characterization of Weld Residual Stresses on a Full-Diameter SNF Interim Storage Canister Mockup.

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    This report documents the mockup specifications and manufacturing processes; the initial cutting of the mockup into three cylindrical pieces for testing and the measured strain changes that occurred during the cutting process; and the planned weld residual stress characterization activities and the status of those activities.

  20. ASSEMBLY TRANSFER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    B. Gorpani

    2000-06-26

    The Assembly Transfer System (ATS) receives, cools, and opens rail and truck transportation casks from the Carrier/Cask Handling System (CCHS). The system unloads transportation casks consisting of bare Spent Nuclear Fuel (SNF) assemblies, single element canisters, and Dual Purpose Canisters (DPCs). For casks containing DPCs, the system opens the DPCs and unloads the SNF. The system stages the assemblies, transfer assemblies to and from fuel-blending inventory pools, loads them into Disposal Containers (DCs), temporarily seals and inerts the DC, decontaminates the DC and transfers it to the Disposal Container Handling System. The system also prepares empty casks and DPCs for off-site shipment. Two identical Assembly Transfer System lines are provided in the Waste Handling Building (WHB). Each line operates independently to handle the waste transfer throughput and to support maintenance operations. Each system line primarily consists of wet and dry handling areas. The wet handling area includes a cask transport system, cask and DPC preparation system, and a wet assembly handling system. The basket transport system forms the transition between the wet and dry handling areas. The dry handling area includes the dry assembly handling system, assembly drying system, DC preparation system, and DC transport system. Both the wet and dry handling areas are controlled by the control and tracking system. The system operating sequence begins with moving transportation casks to the cask preparation area. The cask preparation operations consist of cask cavity gas sampling, cask venting, cask cool-down, outer lid removal, and inner shield plug lifting fixture attachment. Casks containing bare SNF (no DPC) are filled with water and placed in the cask unloading pool. The inner shield plugs are removed underwater. For casks containing a DPC, the cask lid(s) is removed, and the DPC is penetrated, sampled, vented, and cooled. A DPC lifting fixture is attached and the cask is placed

  1. Fire Hazard Analysis for the Cold Vacuum Drying facility (CVD) Facility

    CERN Document Server

    Singh, G

    2000-01-01

    The CVDF is a nonreactor nuclear facility that will process the Spent Nuclear Fuels (SNF) presently stored in the 105-KE and 105-KW SNF storage basins. Multi-canister overpacks (MCOs) will be loaded (filled) with K Basin fuel transported to the CVDF. The MCOs will be processed at the CVDF to remove free water from the fuel cells (packages). Following processing at the CVDF, the MCOs will be transported to the CSB for interim storage until a long-term storage solution can be implemented. This operation is expected to start in November 2000. A Fire Hazard Analysis (FHA) is required for all new facilities and all nonreactor nuclear facilities, in accordance with U.S. Department of Energy (DOE) Order 5480.7A, Fire Protection. This FHA has been prepared in accordance with DOE 5480.7A and HNF-PRO-350, Fire Hazard Analysis Requirements. Additionally, requirements or criteria contained in DOE, Richland Operations Office (RL) RL Implementing Directive (RLID) 5480.7, Fire Protection, or other DOE documentation are cite...

  2. Functions of an engineered barrier system for a nuclear waste repository in basalt

    Energy Technology Data Exchange (ETDEWEB)

    Coons, W.E.; Moore, E.L.; Smith, M.J.; Kaser, J.D.

    1980-01-01

    Defined in this document are the functions of components selected for an engineered barrier system for a nuclear waste repository in basalt. The definitions provide a focal point for barrier material research and development by delineating the purpose and operative lifetime of each component of the engineered system. A five-component system (comprised of waste form, canister, buffer, overpack, and tailored backfill) is discussed in terms of effective operation throughout the course of repository history, recognizing that the emplacement environment changes with time. While components of the system are mutually supporting, redundancy is provided by subsystems of physical and chemical barriers which act in concert with the geology to provide a formidable barrier to transport of hazardous materials to the biosphere. The operating philosophy of the conceptual engineered barrier system is clarified by examples pertinent to storage in basalt, and a technical approach to barrier design and material selection is proposed. A method for system validation and qualification is also included which considers performance criteria proposed by external agencies in conjunction with site-specific models and risk assessment to define acceptable levels of system performance.

  3. DPC materials and corrosion environments.

    Energy Technology Data Exchange (ETDEWEB)

    Ilgen, Anastasia Gennadyevna; Bryan, Charles R.; Teich-McGoldrick, Stephanie; Hardin, Ernest

    2014-10-01

    After an exposition of the materials used in DPCs and the factors controlling material corrosion in disposal environments, a survey is given of the corrosion rates, mechanisms, and products for commonly used stainless steels. Research needs are then identified for predicting stability of DPC materials in disposal environments. Stainless steel corrosion rates may be low enough to sustain DPC basket structural integrity for performance periods of as long as 10,000 years, especially in reducing conditions. Uncertainties include basket component design, disposal environment conditions, and the in-package chemical environment including any localized effects from radiolysis. Prospective disposal overpack materials exist for most disposal environments, including both corrosion allowance and corrosion resistant materials. Whereas the behavior of corrosion allowance materials is understood for a wide range of corrosion environments, demonstrating corrosion resistance could be more technically challenging and require environment-specific testing. A preliminary screening of the existing inventory of DPCs and other types of canisters is described, according to the type of closure, whether they can be readily transported, and what types of materials are used in basket construction.

  4. Study on assessment safety of geological disposal of high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    Department of Environmental Safety Research of Japan Atomic Energy Research Institute has conducted the study on safety of geological disposal of high level radioactive waste. The long-term safety of the geological disposal is proposed to be secured by the multi barrier system which consists of engineered and natural barriers. Thus, in order to clarify the performance of the engineered barrier, we have studied on the long-term behaviors of waste forms, canister, overpack, back fill materials. We have developed a new waste form, i.e. ceramic waste form. And in order to clarify the performance of the natural barrier, we have studied on the hydrology, rock properties, geochemistry of actinides, sorption and fixation of radionuclides on and to rocks and/or minerals, alteration of minerals, dispersion behavior of radionuclides. Natural analogue studies and in-situ experiments have also been conducted. According to the methodology for the assessment established, the assessment model has been developed. (J.P.N.).

  5. EEC Round Robin test on the repository system simulation - clay option. Final report EEC research contract No. Fl 1W-0116-l

    Energy Technology Data Exchange (ETDEWEB)

    Cantale, C.; Castelli, S.; Donato, A.; Traverso, D.M.

    1989-06-01

    The objective of the 'Reporting System Simulation Test (RSST) was to define a test method for the evaluation of waste performances in various geological disposal scenarios, with the possibility to simulate various additional materials, like canister, backfilling, overpack and host rock. The test procedure for the clay-option needed to be developed, especially with reference to the solution preparation and Eh (redox potential) measurements. The necessity to work in an argon atmosphere glove box brings in some operational complications, mainly in the case of the Eh measurement, filtering and final volume determination. The reproducibility of the experimental data does not seem to be particularly good, probably due to the difficulties in standardizing the experimental procedure. An effort should be done to improve these aspects. The tightness of the container and the physical stability of the rock/backfill/grill with pins/synthetic interstitial claywater system are excellent. The leaching data reveal that the clay contribution to the final ionic concentration of the solution is not negligible and should be better evaluated. The overall data-set shows that the glass leach rate, as expressed by Boron mass loss, decreases with time.

  6. Estudo de liberação in vitro do filtro solar p-metoxicinamato de octila incluso em lipossoma e β-ciclodextrina

    Directory of Open Access Journals (Sweden)

    E. P. SANTOS

    2009-05-01

    Full Text Available O objetivo deste estudo foi comparar a liberação do p-metoxicinamato de octila (MCO a partir de três formulações em gel. A primeira contendo o MCO livre, a segunda contendo o MCO incluso em β-ciclodextrina (β-CD/MCO e a terceira contendo o MCO incluso em lipossoma (lipossoma/MCO. O estudo de liberação foi realizado em células de difusão do tipo Franz usando membrana artifi cial de acetato de celulose. A concentração de MCO liberada foi determinada por cromatografi a líquida de alta efi ciência (CLAE. Os perfi s de liberação in vitro mostraram que a inclusão do MCO nesses sistemas de liberação reduziu a liberação do MCO para a solução receptora comparando com a formulação de MCO livre. Entre as formulações de β-CD/MCO e lipossoma/MCO, a que liberou menor concentração de MCO para a solução receptora foi a formulação de lipossoma/MCO, mostrando os melhores resultados. Palavras-chave: p-metoxicinamato de octila; lipossoma; ciclodextrina; célula de Franz; liberação in vitro.

  7. Multicopper oxidase-1 orthologs from diverse insect species have ascorbate oxidase activity.

    Science.gov (United States)

    Peng, Zeyu; Dittmer, Neal T; Lang, Minglin; Brummett, Lisa M; Braun, Caroline L; Davis, Lawrence C; Kanost, Michael R; Gorman, Maureen J

    2015-04-01

    Members of the multicopper oxidase (MCO) family of enzymes can be classified by their substrate specificity; for example, ferroxidases oxidize ferrous iron, ascorbate oxidases oxidize ascorbate, and laccases oxidize aromatic substrates such as diphenols. Our previous work on an insect multicopper oxidase, MCO1, suggested that it may function as a ferroxidase. This hypothesis was based on three lines of evidence: RNAi-mediated knock down of Drosophila melanogaster MCO1 (DmMCO1) affects iron homeostasis, DmMCO1 has ferroxidase activity, and DmMCO1 has predicted iron binding residues. In our current study, we expanded our focus to include MCO1 from Anopheles gambiae, Tribolium castaneum, and Manduca sexta. We verified that MCO1 orthologs have similar expression profiles, and that the MCO1 protein is located on the basal surface of cells where it is positioned to oxidize substrates in the hemolymph. In addition, we determined that RNAi-mediated knock down of MCO1 in A. gambiae affects iron homeostasis. To further characterize the enzymatic activity of MCO1 orthologs, we purified recombinant MCO1 from all four insect species and performed kinetic analyses using ferrous iron, ascorbate and two diphenols as substrates. We found that all of the MCO1 orthologs are much better at oxidizing ascorbate than they are at oxidizing ferrous iron or diphenols. This result is surprising because ascorbate oxidases are thought to be specific to plants and fungi. An analysis of three predicted iron binding residues in DmMCO1 revealed that they are not required for ferroxidase or laccase activity, but two of the residues (His374 and Asp380) influence oxidation of ascorbate. These two residues are conserved in MCO1 orthologs from insects and crustaceans; therefore, they are likely to be important for MCO1 function. The results of this study suggest that MCO1 orthologs function as ascorbate oxidases and influence iron homeostasis through an unknown mechanism.

  8. Medicaid and Managed Care: Key Data, Trends, and Issues

    Science.gov (United States)

    ... are now in publicly traded plans (Figure 2). Distribution of Medicaid MCO Enrollment by Selected MCO Characteristics ... to managed care for this purpose. Managed care companies are also planning for 2014, positioning themselves to ...

  9. Technology status in support of refined technical baseline for the Spent Nuclear Fuel project. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Puigh, R.J.; Toffer, H.; Heard, F.J.; Irvin, J.J.; Cooper, T.D.

    1995-10-20

    The Spent Nuclear Fuel Project (SNFP) has undertaken technology acquisition activities focused on supporting the technical basis for the removal of the N Reactor fuel from the K Basins to an interim storage facility. The purpose of these technology acquisition activities has been to identify technology issues impacting design or safety approval, to establish the strategy for obtaining the necessary information through either existing project activities, or the assignment of new work. A set of specific path options has been identified for each major action proposed for placing the N Reactor fuel into a ``stabilized`` form for interim storage as part of this refined technical basis. This report summarizes the status of technology information acquisition as it relates to key decisions impacting the selection of specific path options. The following specific categories were chosen to characterize and partition the technology information status: hydride issues and ignition, corrosion, hydrogen generation, drying and conditioning, thermal performance, criticality and materials accountability, canister/fuel particulate behavior, and MCO integrity. This report represents a preliminary assessment of the technology information supporting the SNFP. As our understanding of the N Reactor fuel performance develops the technology information supporting the SNFP will be updated and documented in later revisions to this report. Revision 1 represents the incorporation of peer review comments into the original document. The substantive evolution in our understanding of the technical status for the SNFP (except section 3) since July 1995 have not been incorporated into this revision.

  10. Biocatalytic potential of laccase-like multicopper oxidases from Aspergillus niger

    NARCIS (Netherlands)

    Tamayo Ramos, J.A.; Berkel, van W.J.H.; Graaff, de L.H.

    2012-01-01

    BACKGROUND: Laccase-like multicopper oxidases have been reported in several Aspergillus species but they remain uncharacterized. The biocatalytic potential of the Aspergillus niger fungal pigment multicopper oxidases McoA and McoB and ascomycete laccase McoG was investigated. RESULTS: The laccase-li

  11. Cold Vacuum Drying (CVD) Set Point Determination

    Energy Technology Data Exchange (ETDEWEB)

    PHILIPP, B.L.

    2000-03-21

    The Safety Class Instrumentation and Control (SCIC) system provides active detection and response to process anomalies that, if unmitigated, would result in a safety event. Specifically, actuation of the SCIC system includes two portions. The portion which isolates the MCO and initiates the safety-class helium (SCHe) purge, and the portion which detects and stops excessive heat input to the MCO on high tempered water MCO inlet temperature. For the MCO isolation and purge, the SCIC receives signals from MCO pressure (both positive pressure and vacuum), helium flow rate, bay high temperature switches, seismic trips and time under vacuum trips.

  12. Uranium Pyrophoricity Phenomena and Prediction (FAI/00-39)

    Energy Technology Data Exchange (ETDEWEB)

    PLYS, M.G.

    2000-10-10

    The purpose of this report is to provide a topical reference on the phenomena and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel (SNF) Project with specific applications to SNF Project processes and situations. Spent metallic uranium nuclear fuel is currently stored underwater at the K basins in the Hanford 100 area, and planned processing steps include: (1) At the basins, cleaning and placing fuel elements and scrap into stainless steel multi-canister overpacks (MCOs) holding about 6 MT of fuel apiece; (2) At nearby cold vacuum drying (CVD) stations, draining, vacuum drying, and mechanically sealing the MCOs; (3) Shipping the MCOs to the Canister Storage Building (CSB) on the 200 Area plateau; and (4) Welding shut and placing the MCOs for interim (40 year) dry storage in closed CSB storage tubes cooled by natural air circulation through the surrounding vault. Damaged fuel elements have exposed and corroded fuel surfaces, which can exothermically react with water vapor and oxygen during normal process steps and in off-normal situations, A key process safety concern is the rate of reaction of damaged fuel and the potential for self-sustaining or runaway reactions, also known as uranium fires or fuel ignition. Uranium metal and one of its corrosion products, uranium hydride, are potentially pyrophoric materials. Dangers of pyrophoricity of uranium and its hydride have long been known in the U.S. Department of Energy (Atomic Energy Commission/DOE) complex and will be discussed more below; it is sufficient here to note that there are numerous documented instances of uranium fires during normal operations. The motivation for this work is to place the safety of the present process in proper perspective given past operational experience. Steps in development of such a perspective are: (1) Description of underlying physical causes for runaway reactions, (2) Modeling physical processes to explain runaway reactions, (3) Validation of the method

  13. Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D&D) and to reduce the cost of maintaining the facilities prior to D&D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor`s fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered.

  14. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Ultrasonic imaging of EB weld, theory of harmonic imaging of welds, NDE of cast iron

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, T.; Lingvall, F.; Ping Wu [Uppsala Univ. (Sweden). Dept. of Materials Science

    2001-07-01

    The objective of task presented in the first chapter, ultrasonic imaging of EB weld is to investigate imaging methods capable of improving ultrasonic imaging of defects in EB-welds. Algorithms based on ideas from ultrasonic tomography were examined as the first step. After a concise review of literature in the field of tomography the attention is focused on synthetic focusing and particularly on using linear phased array systems for imaging. Synthetic focusing is a technique where the focusing is performed by software after gathering the ultrasonic data. General principles of synthetic aperture focusing technique (SAFT) - a synthetic focusing technique especially suitable for linear ultrasonic arrays are presented. Problems related to the application of SAFT to ultrasonic transducers with large apertures are identified and the solution is proposed. It appears that when the probe becomes larger (i.e., cannot be regarded as a point source) the ultrasonic pulses that it generates will be smeared by its spatial impulse response (SIR). This impairs the spatial resolution achieved for the finite aperture probes comparing to the point source. Thus, a proper application of synthetic focusing requires taking into account the spatially varying probe's SIR. The SIR has to be calculated (measured) in the interesting points of space and than deconvoluted. A technique for deconvoluting the SIR based on Wiener filter is proposed and illustrated by experimental results. Some preliminary results from immersion testing of copper blocks using the ALLIN system in our lab facility are presented. Nonlinear propagation of plane waves in fluids based on the Burgers equation is investigated in the second chapter. The presented method is basically adopted from the existing literature although some modification has been made to adapt to our situation. The solution has been re-derived and two alternative forms feasible for computer calculation are given and some numerical results are presented. The calculated results show how the harmonics evolve as the plane wave propagates. It should be noted that the work presented here is at its preliminary stage, the goal of the present and future work is to build a simulating tool for material harmonic imaging technology. The theory of phase conjugation is presented and different methods of wave phase conjugation (WPC) are reviewed and characterized in the third chapter. The ability of WPC to self-adaptive focus ultrasonic waves in inhomogeneous media makes it interesting in the application to the inspection of as EB welds. The WPC can be performed either in time or frequency domain. Time domain method, known as time reversal mirrors is reviewed in some detail with focus on its applications to NDT. Frequency domain techniques use nonlinear piezoelectric or magnetic materials. The choice of magneto-acoustic phase conjugation, performed in nonlinear magnetic ceramics as a candidate for the feasibility demonstration is motivated. Details of the preliminary experiment with high frequency NDE application (10 MHz) are presented. NDE methods suitable for the characterization of cast iron are reviewed in the fourth chapter. Two groups of methods that could be used in an industrial environment, those based on ultrasound and on eddy current measurement are presented in some detail. The review is focused on sensing the interaction of elastic waves with the microstructure of cast iron. It is explained how three different features of ultrasound, the sound velocity, the attenuation and the backscattering, can be used for the characterization.

  15. Recombinant Expression and Phenotypic Screening of a Bioactive Cyclotide Against α-Synuclein-Induced Cytotoxicity in Baker's Yeast.

    Science.gov (United States)

    Jagadish, Krishnappa; Gould, Andrew; Borra, Radhika; Majumder, Subhabrata; Mushtaq, Zahid; Shekhtman, Alexander; Camarero, Julio A

    2015-07-13

    We report for the first time the recombinant expression of fully folded bioactive cyclotides inside live yeast cells by using intracellular protein trans-splicing in combination with a highly efficient split-intein. This approach was successfully used to produce the naturally occurring cyclotide MCoTI-I and the engineered bioactive cyclotide MCoCP4. Cyclotide MCoCP4 was shown to reduce the toxicity of human α-synuclein in live yeast cells. Cyclotide MCoCP4 was selected by phenotypic screening from cells transformed with a mixture of plasmids encoding MCoCP4 and inactive cyclotide MCoTI-I in a ratio of 1:5×10(4). This demonstrates the potential for using yeast to perform phenotypic screening of genetically encoded cyclotide-based libraries in eukaryotic cells.

  16. Aqueous Corrosion Rates for Waste Package Materials

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. NOTES. A Course Relating Agronomy and Science to Society.

    Science.gov (United States)

    McIntosh, Marla S.

    1993-01-01

    Describes a course designed to teach the relationship between science, agronomy, and society. Includes course and class description, course content, and evaluation of the course. (11 references) (MCO)

  18. Waste disposal package

    Science.gov (United States)

    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.

  19. Corrosion resistant storage container for radioactive material

    Science.gov (United States)

    Schweitzer, Donald G.; Davis, Mary S.

    1990-01-01

    A corrosion resistant long-term storage container for isolating radioactive waste material in a repository. The container is formed of a plurality of sealed corrosion resistant canisters of different relative sizes, with the smaller canisters housed within the larger canisters, and with spacer means disposed between judxtaposed pairs of canisters to maintain a predetermined spacing between each of the canisters. The combination of the plural surfaces of the canisters and the associated spacer means is effective to make the container capable of resisting corrosion, and thereby of preventing waste material from leaking from the innermost canister into the ambient atmosphere.

  20. Dicty_cDB: Contig-U03331-1 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available raphidium curvatum mitochondr... 46 4e-04 DQ026513_4( DQ026513 |pid:none) Candida orthopsilosis strain MCO47...... 46 4e-04 AY962590_4( AY962590 |pid:none) Candida orthopsilosis strain MCO45... 46 4e-04 AJ841288_1( AJ84

  1. Dicty_cDB: FC-IC0905 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available hon... 39 0.057 DQ026513_4( DQ026513 |pid:none) Candida orthopsilosis strain MCO4...7... 39 0.057 AY962590_4( AY962590 |pid:none) Candida orthopsilosis strain MCO45... 39 0.057 AJ249985_2( AJ2

  2. Topical Moltkia coerulea hydroethanolic extract accelerates the repair of excision wound in a rat model

    Institute of Scientific and Technical Information of China (English)

    Mohammad Reza Farahpour; Aydin Dilmaghanian; Maisam Faridy; Esmaeil Karashi

    2016-01-01

    Purpose:To evaluate the effect of a hydroethanolic extract of Moltkia coerulea ointment (MCO) on the healing of excision wound in a rat model.Methods:Circular surgical full thickness excision wound,with 314 mm2 size,was induced in the anterior-dorsal side of each rat.Three different doses of MCO (1%,3% and 6%) were administrated.On Day 3,7,14 and 21,the tissue was sampled and immune cells,fibroblasts and fibrocytes distribution per one mm2 of wound area,collagen density and re-epithelialization were analyzed.Moreover,the total flavnoid,phenols and anti-oxidant potential of the MCO were evaluated.Ultimately,the percentage of wound contraction in different groups was compared with each other.Results:Hydroethanolic extract of MCO significantly (p < 0.05) increased wound contraction percentage.The animals in medium and high dose MCO-treated groups exhibited remarkably (p < 0.05) higher fibroblast and fibrocyte distribution and significantly (p < 0.05) lower immune cells infiltration.On Day 7 after injury,MCO up-regulated neovascularization in a dose-dependent way.Conclusion:Our data showed that MCO shortened the inflammation phase by provoking the fibroblast proliferation.Moreover,MCO promoted the healing process by up-regulating the angiogenesis and provoking the structural cells proliferation as well as increasing the collagen synthesis,cross-linking,and deposition.

  3. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-07-25

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on {approx}0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford ''New Production Reactor'', later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory's (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of

  4. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-09-15

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on ~0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford “New Production Reactor,” later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory’s (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium

  5. Minerals as natural analogues for crystalline nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Giere, R. [Purdue University West Lafayette, Earth and Atmospheric Sciences (United States)

    2000-07-01

    Between the mining of uranium ore (mostly as uraninite) and the final disposal of nuclear waste, there are many processes and steps which together comprise the nuclear fuel cycle. Radioactive waste will be generated as long as nuclear reactors are in operation, but it is also produced by other means, e.g., during certain medical, scientific and industrial procedures. The most dangerous wastes are those resulting from the reprocessing of spent nuclear fuel and from some processes in the production and dismantling of nuclear weapons. A large part of this highly radioactive waste is present as a liquid and thus, its safe isolation from the biosphere requires immobilization of the radionuclides in a durable matrix (waste form). This is a solid which must be resistant to heat, radiation and corrosion over a geologic time scale. Three main categories of waste forms have been developed for the immobilization of radioactive waste, namely glasses, crystalline and multibarrier waste forms. One of the key properties of a nuclear waste form is its chemical durability (or resistance to corrosion), because the waste form represents the primary barrier to radionuclide release. The sciences of mineralogy and petrology have both contributed significantly to the development, characterization and performance assessment of such waste forms. The most important goal of safe nuclear waste disposal is to ensure that practically no radioactive materials reach the biosphere and, ultimately, human beings. Therefore, the design of final repositories is based on an approach that places several obstacles, or barriers, between waste and biosphere, whereby each barrier has a specific role in preventing or delaying migration of radioactive material. This multibarrier concept is different for each type of waste but, for the option of geological disposal, it generally comprises the following five barriers: (1) waste form (contains the actual waste); (2) canister (surrounds waste form; composed of a

  6. Cold Vacuum Drying (CVD) Set Point Determination

    Energy Technology Data Exchange (ETDEWEB)

    PHILIPP, B.L.

    2000-01-12

    This document provides the calculations used to determine the error of safety class signals used for the CVD process These errors are used with the Parameter limits to arrive at the initial set point. The Safety Class Instrumentation and Control (SCIC) system provides active detection and response to process anomalies that, if unmitigated would result in a safety event. Specifically actuation of the SCIC system includes two portions. The portion which isolates the MCO and initiates the safety-class helium (SCHe) purge, and the portion which detects and stops excessive heat input to the MCO on high tempered water MCO inlet temperature. For the MCO isolation and purge the SCIC receives signals from MCO pressure (both positive pressure and vacuum) helium flow rate, bay high temperature switches, seismic trips and time under vacuum trips.

  7. Preliminary Thermal Modeling of HI-STORM 100 Storage Modules at Diablo Canyon Power Plant ISFSI

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Adkins, Harold E.

    2014-04-17

    Thermal analysis is being undertaken at Pacific Northwest National Laboratory (PNNL) in support of inspections of selected storage modules at various locations around the United States, as part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development. This report documents pre-inspection predictions of temperatures for two modules at the Diablo Canyon Power Plant ISFSI identified as candidates for inspection. These are HI-STORM 100 modules of a site-specific design for storing PWR 17x17 fuel in MPC-32 canisters. The temperature predictions reported in this document were obtained with detailed COBRA-SFS models of these storage systems, with the following boundary conditions and assumptions. • storage module overpack configuration based on FSAR documentation of HI-STORM100S-218, Version B; due to unavailability of site-specific design data for Diablo Canyon ISFSI modules • Individual assembly and total decay heat loadings for each canister, based on at-loading values provided by PG&E, “aged” to time of inspection using ORIGEN modeling o Special Note: there is an inherent conservatism of unquantified magnitude – informally estimated as up to approximately 20% -- in the utility-supplied values for at-loading assembly decay heat values • Axial decay heat distributions based on a bounding generic profile for PWR fuel. • Axial location of beginning of fuel assumed same as WE 17x17 OFA fuel, due to unavailability of specific data for WE17x17 STD and WE 17x17 Vantage 5 fuel designs • Ambient conditions of still air at 50°F (10°C) assumed for base-case evaluations o Wind conditions at the Diablo Canyon site are unquantified, due to unavailability of site meteorological data o additional still-air evaluations performed at 70°F (21°C), 60°F (16°C), and 40°F (4°C), to cover a range of possible conditions at the time of the inspection. (Calculations were also performed at

  8. Determination of VOCs in In-door Smoking Air by GC/MS with Canister Sampling%SUMMA罐采样-GC/MS法测定吸烟室内空气中挥发性有机物

    Institute of Scientific and Technical Information of China (English)

    杨丽莉; 王美飞; 胡恩宇

    2011-01-01

    采用空气预浓缩与气相色谱/质谱联用技术对空气中59种痕量挥发性有机化合物进行定性与定量分析,应用研究的技术对吸烟室烟草空气中的挥发性有机物成分定性解析,对59种常见挥发性有机污染物定量检测.室内环境烟草空气中检出多种挥发性有机污染物,主要有烯烃、烷烃、苯系物等有害成分,不仅对被动吸烟人群造成危害,同时也影响大气环境质量.%A determination method of 59 volatile organic compounds ( VOCs) in ambient air by air pre-con-centration and gas chromatography-mass spectrometry has been studied. VOCs in air of smoking room was qualitatively analyzed and 59 VOCs was quantitatively detected. Some VOCs in the air were hazardous pollutants such as olefins, alkanes, and aromatic hydrocarbons. These compounds not only harmed to passive smoking people but also affected the atmospheric environmental quality.

  9. 同心筒发射装置导弹燃气流热效应数值模拟%Numerical Simulation of Thermal Effect of Missile Combustion-gas Flow for Concentric Canister Launcher

    Institute of Scientific and Technical Information of China (English)

    蔺翠郎; 毕世华

    2008-01-01

    针对某型舰载导弹同心筒发射装置,采用动态网格技术,建立了导弹发射过程中燃气流与发射筒之间的流固耦合换热模型.模型中考虑了导弹在发射筒内运动引起的燃气流场边界的变化.通过对燃气流非稳态传热的数值模拟,得到了发射筒内燃气流及筒壁的温度分布规律,为发射筒的热强度设计提供依据.

  10. 微重力下高温固液相变蓄热容器内空穴分布%INITIAL STUDY OF VOID FORMATION OF HIGH TEMPERATURE SOLID-LIQUID PHASE CHANGE THERMAL ENERGY STORAGE CANISTER IN MICROGRAVITY

    Institute of Scientific and Technical Information of China (English)

    邢玉明; 崔海亭; 袁修干; 粟卫芳

    2003-01-01

    高温固液相变蓄热容器是空间太阳能动力装置吸热-储热器的关键部件.作为相变材料(PCM)的氟盐在凝固时体积收缩很大,从而在PCM容器内形成空穴.空穴的存在增大了传热热阻,还可能使PCM容器产生"热斑"和"热松脱"现象.该文建立了微重力下基于焓法形式的二维数学模型和一个改进的空穴模型,提出了计算相变过程中空穴体积变化及空穴调整的算法.预测了PCM容器在一个轨道周期内的空穴分布.计算结果有助于解决PCM容器的"热斑"和"热松脱"问题.

  11. Thermal Effects of Structural Parameters of PCM Canister on Heat Receiver under a Certain Working Condition%PCM容器的结构参数对蓄热单元热性能的影响

    Institute of Scientific and Technical Information of China (English)

    应歌; 杜朝辉; 王平阳; 钱中

    2006-01-01

    适合空间设备的主要电力来源就是太阳能热动力发电系统.为了深入考察蓄热容器(PCM容器)的结构参数对空间太阳能热动力发电系统的关键部件之一,即吸热蓄热器热性能的影响,建立了PCM容器的二维热分析模型,并在两种工作参数条件下对不同径向高度的PCM容器进行了数值计算,结果表明PCM容器的外径对吸热蓄热器热性能具有重要影响.研究结果可为提高PCM容器的功率质量比提供参考依据.

  12. Alternative technical summary report for immobilized disposition in deep boreholes: Immobilized disposal of plutonium in coated ceramic pellets in grout without canisters, Version 4.0. Fissile materials disposition program

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.

    1996-08-23

    This paper summarizes and compares the immobilized and direct borehole disposition alternatives previously presented in the alternative technical summary. The important design concepts, facility features and operational procedures are first briefly described. This is followed by a discussion of the issues that affect the evaluation of each alternative against the programmatic assessment criteria that have been established for selecting the preferred alternatives for plutonium disposition.

  13. Fissile Material Disposition Program: Deep Borehole Disposal Facility PEIS data input report for direct disposal. Direct disposal of plutonium metal/plutonium dioxide in compound metal canisters. Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-15

    The US Department of Energy (DOE) is examining options for disposing of excess weapons-usable nuclear materials [principally plutonium (Pu) and highly enriched uranium (HEU)] in a form or condition that is substantially and inherently more difficult to recover and reuse in weapons production. This report is the data input report for the Programmatic Environmental Impact Statement (PEIS). The PEIS examines the environmental, safety, and health impacts of implementing each disposition alternative on land use, facility operations, and site infrastructure; air quality and noise; water, geology, and soils; biotic, cultural, and paleontological resources; socioeconomics; human health; normal operations and facility accidents; waste management; and transportation. This data report is prepared to assist in estimating the environmental effects associated with the construction and operation of a Deep Borehole Disposal Facility, an alternative currently included in the PEIS. The facility projects under consideration are, not site specific. This report therefore concentrates on environmental, safety, and health impacts at a generic site appropriate for siting a Deep Borehole Disposal Facility.

  14. Fissile Material Disposition Program: Deep borehole disposal Facility PEIS date input report for immobilized disposal. Immobilized disposal of plutonium in coated ceramic pellets in grout with canisters. Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-15

    Following President Clinton`s Non-Proliferation Initiative, launched in September, 1993, an Interagency Working Group (IWG) was established to conduct a comprehensive review of the options for the disposition of weapons-usable fissile materials from nuclear weapons dismantlement activities in the United States and the former Soviet Union. The IWG review process will consider technical, nonproliferation, environmental budgetary, and economic considerations in the disposal of plutonium. The IWG is co-chaired by the White House Office of Science and Technology Policy and the National Security Council. The Department of Energy (DOE) is directly responsible for the management, storage, and disposition of all weapons-usable fissile material. The Department of Energy has been directed to prepare a comprehensive review of long-term options for Surplus Fissile Material (SFM) disposition, taking into account technical, nonproliferation, environmental, budgetary, and economic considerations.

  15. Development of the Virtual Instrument for the Active Charcoal Canister Used in Vehicle with Petrol Engine%汽油车用活性炭罐性能测试系统的虚拟仪器

    Institute of Scientific and Technical Information of China (English)

    邴冰; 刘长良; 许善珍; 耿松亮

    2007-01-01

    为了更好地执行国家标准HBC 32-2004《环境保护产品认定技术要求汽油车燃油蒸发污染物控制系统》,利用虚拟仪器开发平台LabVIEW7.1研制一套汽油车用活性炭罐性能测试系统,实现了系统的自动测控.

  16. Evaluation of design and operation basis of the smear test station

    Energy Technology Data Exchange (ETDEWEB)

    Hutsell, D.J.

    2000-02-29

    The purpose of the WTC-STS is to provide final verification that the external canister surface is free of transferable contamination before transporting the canister to the Glass Waste Storage Building for onsite storage.

  17. The machine conception of the organism in development and evolution: a critical analysis.

    Science.gov (United States)

    Nicholson, Daniel J

    2014-12-01

    This article critically examines one of the most prevalent metaphors in contemporary biology, namely the machine conception of the organism (MCO). Although the fundamental differences between organisms and machines make the MCO an inadequate metaphor for conceptualizing living systems, many biologists and philosophers continue to draw upon the MCO or tacitly accept it as the standard model of the organism. The analysis presented here focuses on the specific difficulties that arise when the MCO is invoked in the contexts of development and evolution. In developmental biology the MCO underlies a logically incoherent model of ontogeny, the genetic program, which serves to legitimate three problematic theses about development: genetic animism, neo-preformationism, and developmental computability. In evolutionary biology the MCO is responsible for grounding unwarranted theoretical appeals to the concept of design as well as to the interpretation of natural selection as an engineer, which promote a distorted understanding of the process and products of evolutionary change. Overall, it is argued that, despite its heuristic value, the MCO today is impeding rather than enabling further progress in our comprehension of living systems.

  18. 3D conformal planning using low segment multi-criteria IMRT optimization

    CERN Document Server

    Khan, Fazal

    2014-01-01

    Purpose: To evaluate automated multicriteria optimization (MCO)-- designed for intensity modulated radiation therapy (IMRT), but invoked with limited segmentation -- to efficiently produce high quality 3D conformal treatment (3D-CRT) plans. Methods: Ten patients previously planned with 3D-CRT were replanned with a low-segment inverse multicriteria optimized technique. The MCO-3D plans used the same number of beams, beam geometry and machine parameters of the corresponding 3D plans, but were limited to an energy of 6 MV. The MCO-3D plans were optimized using a fluence-based MCO IMRT algorithm and then, after MCO navigation, segmented with a low number of segments. The 3D and MCO-3D plans were compared by evaluating mean doses to individual organs at risk (OARs), mean doses to combined OARs, homogeneity indexes (HI), monitor units (MUs), physician preference, and qualitative assessments of planning time and plan customizability. Results: The MCO-3D plans significantly reduced the OAR mean doses and monitor unit...

  19. Cold Vacuum Drying (CVD) Set Point Determination

    Energy Technology Data Exchange (ETDEWEB)

    PHILIPP, B.L.

    2000-09-04

    The Safety Class Instrumentation and Control (SCIC) system provides active detection and response to process anomalies that, if unmitigated, would result in a safety event. Specifically, actuation of the SCIC system includes two portions. The portion which isolates the MCO and initiates the safety-class helium (SCHe) purge, and the portion which detects and stops excessive heat input to the MCO annulus on high Tempered Water (TW) inlet temperature. For the MCO isolation and purge, the SCIC receives MCO pressure (both positive pressure and vacuum), helium flow rate, bay high temperature switch status, seismic trip status, and time-under-vacuum trips signals. The SCIC system will isolate the MCO and start an SCHe system purge if any of the following occur. (1) Isolation and purge from one of the SCHe ''isolation'' and ''purge'' buttons is manually initiated (administratively controlled). (2) The first vacuum cycle exceeds 8 hours at vacuum, or any subsequent vacuum cycle exceeds 4 hours at vacuum without re-pressurizing the MCO for a minimum of 4 hours. (This is referred to as the 8/4/4 requirement and provides thermal equilibrium within the MCO.) (3) MCO is below atmospheric pressure and the helium flow is below the minimum required to keep hydrogen less than 4% by volume. (When MCO pressure is below 12 torr there is insufficient hydrogen to exceed the 4% level and therefore no purge is required. A five minute time delay on low flow allows flow to be stopped in order to reach < 12 torr.) (4) The duration for the transition from above atmospheric pressure to vacuum (time to reach pressure below -11.7 psig [{approx}155 torr]) exceeds 5 minutes. (5) The duration for the transition from vacuum (below -11.1 psig [{approx}185 torr]) back to pressure [greater than 0.5 psig] exceeds 5 minutes. (6) MCO reaches a vacuum state (<0.5 psig) without an adequate, verified purge volume. (The MCO must be maintained above atmospheric pressure

  20. Population Education Programme in the 90s in India.

    Science.gov (United States)

    Muley, D. S.

    1991-01-01

    Presents a projection of the plans and requirements for modification of the national population education program in India. Outlines various strategies for attainment of the main objective to institutionalize population education in all national educational processes. (MCO)

  1. Population Education in China: Now and into the Future.

    Science.gov (United States)

    Taijin, Zhang

    1991-01-01

    Presents a comprehensive plan for the expansion of China's population education program. Evaluations which demonstrate the success of the program are the foundation for the plans that address new trends and requirements for population management. (MCO)

  2. 42 CFR 438.806 - Prior approval.

    Science.gov (United States)

    2010-10-01

    ...— (1) The Regional Office has confirmed that the contractor meets the definition of an MCO or is one of...) For subsequent years, the amount is increased by the percentage increase in the consumer price...

  3. Content modification attacks on consensus seeking multi-agent system with double-integrator dynamics.

    Science.gov (United States)

    Dong, Yimeng; Gupta, Nirupam; Chopra, Nikhil

    2016-11-01

    In this paper, vulnerability of a distributed consensus seeking multi-agent system (MAS) with double-integrator dynamics against edge-bound content modification cyber attacks is studied. In particular, we define a specific edge-bound content modification cyber attack called malignant content modification attack (MCoMA), which results in unbounded growth of an appropriately defined group disagreement vector. Properties of MCoMA are utilized to design detection and mitigation algorithms so as to impart resilience in the considered MAS against MCoMA. Additionally, the proposed detection mechanism is extended to detect the general edge-bound content modification attacks (not just MCoMA). Finally, the efficacies of the proposed results are illustrated through numerical simulations.

  4. Ferns.

    Science.gov (United States)

    Russell, Helen Ross

    1991-01-01

    Contains several study methods using ferns. Includes exercises on fern propagation, gametophytes and fern hybrids, fern collection, microscope use, asexual reproduction, and fern photography. Background information describes identification techniques and the alternation of generations phenomenon. (MCO)

  5. Visualizing the Impacts of Deforestation.

    Science.gov (United States)

    Fortner, Rosanne W.

    1992-01-01

    Presents two activities with investigation procedures to aid students in examining the extent and impact of biomass burning and deforestation in Brazil as an example of the global problem. Provides background information, tables, and diagrams. (five references) (MCO)

  6. Soil and Litter Animals.

    Science.gov (United States)

    Lippert, George

    1991-01-01

    A lesson plan for soil study utilizes the Tullgren extraction method to illustrate biological concepts. It includes background information, equipment, collection techniques, activities, and references for identification guides about soil fauna. (MCO)

  7. What Light through Yonder Window Breaks?--The Greenhouse Effect Revisited.

    Science.gov (United States)

    Bohren, Craig F.

    1992-01-01

    Presents three experiments exploring aspects of the greenhouse effect. Topics and discussion includes radiation in energy transfer, emissivity and absorptivity, the irrelevance of reflectivity, a digression on insulators and convection, climate change, and radiative energy balance. (MCO)

  8. Composting: Great Rotten Idea.

    Science.gov (United States)

    Chemecology, 1992

    1992-01-01

    To help students investigate both the advantages and disadvantages of composting, various activities are presented dealing with the definitions and the applications of the concepts of recyclable and biodegradable. (MCO)

  9. Make Your Own Solar Panel.

    Science.gov (United States)

    Suzuki, David

    1992-01-01

    Presents an activity in which students make a simulated solar panel to learn about the principles behind energy production using solar panels. Provides information about how solar panels function to produce energy. (MCO)

  10. Biosynthesis of a Fully Functional Cyclotide inside Living Bacterial Cells

    Energy Technology Data Exchange (ETDEWEB)

    Camarero, J A; Kimura, R H; Woo, Y; Cantor, J; Shekhtman, A

    2007-04-05

    The cyclotide MCoTI-II is a powerful trypsin inhibitor recently isolated from the seeds of Momordica cochinchinensis, a plant member of cucurbitaceae family. We report for the first time the in vivo biosynthesis of natively-folded MCoTI-II inside live E. coli cells. Our biomimetic approach involves the intracellular backbone cyclization of a linear cyclotide-intein fusion precursor mediated by a modified protein splicing domain. The cyclized peptide then spontaneously folds into its native conformation. The use of genetically engineered E. coli cells containing mutations in the glutathione and thioredoxin reductase genes considerably improves the production of folded MCoTI-II in vivo. Biochemical and structural characterization of the recombinant MCoTI-II confirmed its identity. Biosynthetic access to correctly-folded cyclotides allows the possibility of generating cell-based combinatorial libraries that can be screened inside living cells for their ability to modulate or inhibit cellular processes.

  11. Synthesis and infrared spectra of alkaline earth metal carbonates formed by the reaction of metal salts with urea at high temperature

    Indian Academy of Sciences (India)

    S M Teleb; D El-Sayed Nassr; E M Nour

    2004-12-01

    The metal carbonate, MCO3 (M = Ca, Sr and Ba), was synthesized by a novel method of reacting aqueous solution of each of Ca2+, Sr2+ and Ba2+ salts with urea at high temperature, ∼ 80°C. The reaction products were characterized through elemental analysis and infrared spectra. The infrared spectra of the products are the same as those of the corresponding commercially obtained carbonates. A general reaction describing the formation of MCO3 is proposed.

  12. Maximizing dosimetric benefits of IMRT in the treatment of localized prostate cancer through multicriteria optimization planning

    Energy Technology Data Exchange (ETDEWEB)

    Wala, Jeremiah; Craft, David [Harvard Medical School, Boston, MA (United States); Department of Radiation Oncology, Massachusetts General Hospital, Boston, MA (United States); Paly, Jon [Department of Radiation Oncology, Massachusetts General Hospital, Boston, MA (United States); Zietman, Anthony [Harvard Medical School, Boston, MA (United States); Department of Radiation Oncology, Massachusetts General Hospital, Boston, MA (United States); Efstathiou, Jason, E-mail: jefstathiou@partners.org [Harvard Medical School, Boston, MA (United States); Department of Radiation Oncology, Massachusetts General Hospital, Boston, MA (United States)

    2013-10-01

    We examine the quality of plans created using multicriteria optimization (MCO) treatment planning in intensity-modulated radiation therapy (IMRT) in treatment of localized prostate cancer. Nine random cases of patients receiving IMRT to the prostate were selected. Each case was associated with a clinically approved plan created using Corvus. The cases were replanned using MCO-based planning in RayStation. Dose-volume histogram data from both planning systems were presented to 2 radiation oncologists in a blinded evaluation, and were compared at a number of dose-volume points. Both physicians rated all 9 MCO plans as superior to the clinically approved plans (p<10{sup −5}). Target coverage was equivalent (p = 0.81). Maximum doses to the prostate and bladder and the V50 and V70 to the anterior rectum were reduced in all MCO plans (p<0.05). Treatment planning time with MCO took approximately 60 minutes per case. MCO-based planning for prostate IMRT is efficient and produces high-quality plans with good target homogeneity and sparing of the anterior rectum, bladder, and femoral heads, without sacrificing target coverage.

  13. 中国高放废物处置库缓冲材料物理性能%PHYSICAL PROPERTY OF CHINA'S BUFFER MATERIAL FOR HIGH-LEVEL RADIOACTIVE WASTE REPOSITORIES

    Institute of Scientific and Technical Information of China (English)

    温志坚

    2006-01-01

    The deep geological disposal is regarded as the most reasonable and effective way to safely dispose high-level radioactive wastes(HLW) in the world. The conceptual model of HLW geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineered barrier system including the vitrified HLW, canister, overpack and buffer/backfill material. The bentonite is selected as base material of the buffer/backfill material in HLW repositories,due to the very low permeability and excellent retardation of nuclides from migration,etc. GMZ deposit is selected as the candidate supplier for buffer material of HLW repositories in China. Since 2000,systematic study was conducted on GMZ-1 that is Na-bentonite produced from GMZ deposit and selected as reference material for Chinese buffer material study. The mineral composition,basic parameters of GMZ-1 bentonite and thermal conductivity,hydraulic conductivity,unconfined compression strength as function of dry density and water content are presented. The swelling stress of GMZ-1 bentonite as function of dry density is also reported. GMZ-1 bentonite is characterized by high content of montmorillonite(about 75%) and less impurities. The adequacy understanding of property and long-term behavior in deep geological condition of GMZ-1 is essential to safe dispose the high-level radioactive wastes in China.%深地质处置被国际上公认为处置高放废物的最有效可行的方法.中国深地质处置的概念模型采用多重工程屏障系统(包括废物固化体、废物容器、外包装、缓冲/回填材料)和适宜的地质围岩地质体共同作用来确保高放废物与生物圈的安全隔离.膨润土由于具有极低的渗透性和优良的核素吸附等性能而被国际上选作缓冲材料的基础材料.经过全国筛选,高庙子膨润土矿床被选作我国缓冲材料供应基地.从2000年起,对产自该矿床的钠基膨润土GMZ-1开始了

  14. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  15. Assaying Benefits of Poly[styrene-4-(trimethylammonium)methyl Triiodide] in Respiratory Protection Devices

    Science.gov (United States)

    2009-12-01

    Bronze—a clip-on device containing a PSTI-coated nonwoven medium that attached to the front face of a commercial off-the-shelf (COTS) canister...Silver—a fabricated plastic canister that placed a PSTI-coated nonwoven layer in front of the components of the COTS canister) • Gold— a redesigned...humidity. Accommodation of the mouse model was a necessary aspect of the design and construction processes. This system will be used in the

  16. Visual examinations of K west fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

    Over 250 fuel assemblies stored in sealed canisters in the K West Basin were extracted and visually examined for damage. Substantial damage was expected based on high cesium levels previously measured in water samples taken from these canisters. About 11% of the inner elements and 45% of the outer elements were found to be failed in these examinations. Canisters that had cesium levels of I curie or more generally had multiple instances of major fuel damage.

  17. In-Place Filter Tester Instrument for Nuclear Material Containers.

    Science.gov (United States)

    Brown, Austin D; Moore, Murray E; Runnels, Joel T; Reeves, Kirk

    2016-05-01

    A portable instrument was developed to determine filter clogging and container leakage of in-place nuclear material storage canisters. This paper describes the development of an in-place filter tester for determining the "as found" condition of unopened canisters. The U.S. Department of Energy uses several thousand canisters for nuclear material storage, and air filters in the canister lids allow gases to escape while maintaining an equilibrated pressure without release of radioactive contamination. Diagnosing the filter condition and canister integrity is important for ensuring worker and public safety. Customized canister interfaces were developed for suction clamping (during tests) to two of the canister types in use at Los Alamos National Laboratory. Experimental leakage scenarios included: O-rings fouled with dust, cracked O-rings, and loose canister lids. The prototype tester has a measurement range for air leakage rates from 8.2 × 10 mL s up to 3.0 × 10 mL s. This is sufficient to measure a leak rate of 3.4 × 10 mL s, which is the Los Alamos helium leak criterion for post-drop tested canisters. The In-Place-Filter-Tester cannot measure to the lower value of the helium leak criterion for pre-drop tested canisters (1.0 × 10 mL s). However, helium leak testing requires canister disassembly, while the new in-place filter tester is able to assess the assembled condition of as-found and in-situ canisters.

  18. 327 SNF fuel return to K-Basin quality process plan

    Energy Technology Data Exchange (ETDEWEB)

    Ham, J.E.

    1998-09-22

    The B and W Hanford Company`s (BWHC) 327 Facility, in the 300 Area of the Hanford Site, contains Spent Nuclear Fuel (SNF) single fuel element canisters (SFEC) and fuel remnant canisters (FRC) which are to be returned to K-Basin. Seven shipments of up to six fuel canisters will be loaded into the CNS 1-13G Cask and transported to 105-KE.

  19. Multicriteria optimization informed VMAT planning.

    Science.gov (United States)

    Chen, Huixiao; Craft, David L; Gierga, David P

    2014-01-01

    We developed a patient-specific volumetric-modulated arc therapy (VMAT) optimization procedure using dose-volume histogram (DVH) information from multicriteria optimization (MCO) of intensity-modulated radiotherapy (IMRT) plans. The study included 10 patients with prostate cancer undergoing standard fractionation treatment, 10 patients with prostate cancer undergoing hypofractionation treatment, and 5 patients with head/neck cancer. MCO-IMRT plans using 20 and 7 treatment fields were generated for each patient on the RayStation treatment planning system (clinical version 2.5, RaySearch Laboratories, Stockholm, Sweden). The resulting DVH of the 20-field MCO-IMRT plan for each patient was used as the reference DVH, and the extracted point values of the resulting DVH of the MCO-IMRT plan were used as objectives and constraints for VMAT optimization. Weights of objectives or constraints of VMAT optimization or both were further tuned to generate the best match with the reference DVH of the MCO-IMRT plan. The final optimal VMAT plan quality was evaluated by comparison with MCO-IMRT plans based on homogeneity index, conformity number of planning target volume, and organ at risk sparing. The influence of gantry spacing, arc number, and delivery time on VMAT plan quality for different tumor sites was also evaluated. The resulting VMAT plan quality essentially matched the 20-field MCO-IMRT plan but with a shorter delivery time and less monitor units. VMAT plan quality of head/neck cancer cases improved using dual arcs whereas prostate cases did not. VMAT plan quality was improved by fine gantry spacing of 2 for the head/neck cancer cases and the hypofractionation-treated prostate cancer cases but not for the standard fractionation-treated prostate cancer cases. MCO-informed VMAT optimization is a useful and valuable way to generate patient-specific optimal VMAT plans, though modification of the weights of objectives or constraints extracted from resulting DVH of MCO-IMRT or

  20. 49 CFR 175.702 - Separation distance requirements for packages containing Class 7 (radioactive) materials in cargo...

    Science.gov (United States)

    2010-10-01

    ... containing Class 7 (radioactive) materials in cargo aircraft. 175.702 Section 175.702 Transportation Other... (radioactive) materials in cargo aircraft. (a) No person may carry in a cargo aircraft any package required by... separation distance between the surfaces of the radioactive materials packages, overpacks or...

  1. New packaging specifications prompt OC shelf life study.

    Science.gov (United States)

    1983-12-01

    when a major donor of contraceptives changed the packaging specifications of norethindrone/mestranol oral contraceptives (OCs), the change raised some questions as to product shelf life. The new specifications called for the OCs to be packaged in bulk with 100 blister packs/aluminum/plastic overpack. These revised packaging specifications would lead to removal of the blister pack from the protective overpack at an earlier point in the supply chain, and therefore to earlier potential exposure to high temperatures and humidity. However, stability trials conducted by Kabalikat ng Pamilyang Pilipino in Manila have shown no significant deterioration in norethindrone or mestranol content, content uniformity, tablet hardness, disintegration time or blister pack integrity after blister packs without the protective overpack were exposed for 6 months to high humidity and temperatures as high as 60 degrees Celsius. Stability for longer than 6 months remains speculative and should be tested in actual field conditions. Within the limits of the study, however, these results suggest that OCs remain stable with or without the aluminum overpack.

  2. Feasibility of Lateral Emplacement in Very Deep Borehole Disposal of High Level Nuclear Waste

    Science.gov (United States)

    2010-06-01

    including suggestions for reducing this burden, to Washington Headquarters Services , Directorate for Information Operations and Reports, 1215 Jefferson...81 FIGURE 3-32: CANDU GEOLOGIC DISPOSAL OVER-PACK AND...space by Number of Laterals 46 2.6.4 Second Narrowing of the Trade-space Further inspecting the pared-down trade-space results suggested further

  3. Flow Visualization of Forced and Natural Convection in Internal Cavities

    Energy Technology Data Exchange (ETDEWEB)

    John Crepeau; Hugh M. Mcllroy,Jr.; Donald M. McEligot; Keith G. Condie; Glenn McCreery; Randy Clarsean; Robert S. Brodkey; Yann G. Guezennec

    2002-01-31

    The report descries innovative flow visualization techniques, fluid mechanics measurements and computational models of flows in a spent nuclear fuel canister. The flow visualization methods used a fluid that reacted with a metal plate to show how a local reaction affects the surrounding flow. A matched index of refraction facility was used to take mean flow and turbulence measurements within a generic spent nuclear fuel canister. Computational models were also made of the flow in the canister. It was determined that the flow field in the canister was very complex, and modifications may need to be made to ensure that the spent fuel elements are completely passivated.

  4. Visual examinations of K east fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

    Selected fuel elements stored in both ``good fuel`` and ``bad fuel`` canisters in K East Basin were extracted and visually examined full length for damage. Lower end damage in the ``bad fuel`` canisters was found to be more severe than expected based on top end appearances. Lower end damage for the ``good fuel`` canisters, however, was less than expected based on top end observations. Since about half of the fuel in K East Basin is contained in ``good fuel`` canisters based on top end assessments, the fraction of fuel projected to be intact with respect to IPS processing considerations remains at 50% based on these examination results.

  5. MnCo2O4 nanowires anchored on reduced graphene oxide sheets as effective bifunctional catalysts for Li-O2 battery cathodes.

    Science.gov (United States)

    Kim, Jong Guk; Kim, Youngmin; Noh, Yuseong; Kim, Won Bae

    2015-05-22

    A hybrid composite system of MnCo2 O4 nanowires (MCO NWs) anchored on reduced graphene oxide (RGO) nanosheets was prepared as the bifunctional catalyst of a Li-O2 battery cathode. The catalysts can be obtained from the hybridization of one-dimensional MCO NWs and two-dimensional RGO nanosheets. As O2 -cathode catalysts for Li-O2 cells, the MCO@RGO composites showed a high initial discharge capacity (ca. 11092.1 mAh gcarbon (-1) ) with a high rate performance. The Li-O2 cells could run for more than 35 cycles with high reversibility under a limited specific capacity of 1000 mAh gcarbon (-1) with a low potential polarization of 1.36 V, as compared with those of pure Ketjenblack and MCO NWs. The high cycling stability, low potential polarization, and rate capability suggest that the MCO@RGO composites prepared here are promising catalyst candidates for highly reversible Li-O2 battery cathodes.

  6. Fabrication of cubic spinel MnCo2O4 nanoparticles embedded in graphene sheets with their improved lithium-ion and sodium-ion storage properties

    Science.gov (United States)

    Chen, Chang; Liu, Borui; Ru, Qiang; Ma, Shaomeng; An, Bonan; Hou, Xianhua; Hu, Shejun

    2016-09-01

    Cubic Spinel MnCo2O4/graphene sheets (MCO/GS) nanocomposites are synthesized by a facile hydrothermal method with a subsequent annealing process. Nano-sized MnCo2O4 particles are evenly embedded in paper-like graphene sheets, possessing a unique nanoparticles-on-sheets hybrid nanostructure, with particle size around 20-50 nm. Owing to the special nanoparticles-on-sheets structures, MCO/GS nanocomposites have an outstanding electrochemical performance for rechargeable energy storage devices. As an anode material for lithium-ion batteries, MCO/GS electrodes exhibit high reversible discharge capacities (1350.4 mAh g-1 at the initial rate of 100 mA g-1), excellent rate capability (462.1 mAh g-1 at a current rate of 4000 mA g-1) and outstanding cycling performance (584.3 mAh g-1 at 2000 mA g-1 after 250 cycles). Meanwhile, as an anode material for sodium-ion batteries, MCO/GS electrodes also exhibit comparably promising electrochemical characteristics. Greatly improved electrochemical properties can be assigned to the special advantageous nanostructures. Besides, the existence of graphene sheets is beneficial to the transportation of ions/electrons during battery operation. The outstanding electrochemical performance demonstrates that the lithium/sodium storage capability of MCO/GS nanocomposites is highly promising for high-capacity batteries.

  7. Use of engineered soils and other site modifications for low-level radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    The U.S. Nuclear Regulatory Commission requires that low-level radioactive waste (LLW) disposal facilities be designed to minimize contact between waste and infiltrating water through the use of site design features. The purpose of this investigation is to identify engineered barriers and evaluate their ability to enhance the long-term performance of an LLW disposal facility. Previously used barriers such as concrete overpacks, vaults, backfill, and engineered soil covers, are evaluated as well as state-of-the-art barriers, including an engineered sorptive soil layer underlying a facility and an advanced design soil cover incorporating a double-capillary layer. The purpose of this investigation is also to provide information in incorporating or excluding specific engineered barriers as part of new disposal facility designs. Evaluations are performed using performance assessment modeling techniques. A generic reference disposal facility design is used as a baseline for comparing the improvements in long-term performance offered by designs incorporating engineered barriers in generic and humid environments. These evaluations simulate water infiltration through the facility, waste leaching, radionuclide transport through the facility, and decay and ingrowth. They also calculate a maximum (peak annual) dose for each disposal system design. A relative dose reduction factor is calculated for each design evaluated. The results of this investigation are presented for concrete overpacks, concrete vaults, sorptive backfill, sorptive engineered soil underlying the facility, and sloped engineered soil covers using a single-capillary barrier and a double-capillary barrier. Designs using combinations of barriers are also evaluated. These designs include a vault plus overpacks, sorptive backfill plus overpacks, and overpack with vault plus sorptive backfill, underlying sorptive soil, and engineered soil cover.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  9. Reduced malonyl-CoA content in recovery from exercise correlates with improved insulin-stimulated glucose uptake in human skeletal muscle

    DEFF Research Database (Denmark)

    Frøsig, Christian; Roepstorff, Carsten; Brandt, Nina

    2009-01-01

    This study evaluated whether improved insulin-stimulated glucose uptake in recovery from acute exercise coincides with reduced malonyl-CoA (MCoA) content in human muscle. Furthermore, we investigated whether a high-fat diet [65 energy-% (Fat)] would alter the content of MCoA and insulin action...... to be compromised, although to a minor extent, by the Fat diet. Collectively, this study indicates that reduced muscle MCoA content in recovery from exercise may be part of the adaptive response leading to improved insulin action on glucose uptake after exercise in human muscle....... legs during a euglycemic-hyperinsulinemic clamp. Muscle biopsies were obtained in both legs before and after the clamp. Four hours after exercise, insulin-stimulated glucose uptake was improved (approximately 70%, P

  10. Nash equilibrium and multi criterion aerodynamic optimization

    Science.gov (United States)

    Tang, Zhili; Zhang, Lianhe

    2016-06-01

    Game theory and its particular Nash Equilibrium (NE) are gaining importance in solving Multi Criterion Optimization (MCO) in engineering problems over the past decade. The solution of a MCO problem can be viewed as a NE under the concept of competitive games. This paper surveyed/proposed four efficient algorithms for calculating a NE of a MCO problem. Existence and equivalence of the solution are analyzed and proved in the paper based on fixed point theorem. Specific virtual symmetric Nash game is also presented to set up an optimization strategy for single objective optimization problems. Two numerical examples are presented to verify proposed algorithms. One is mathematical functions' optimization to illustrate detailed numerical procedures of algorithms, the other is aerodynamic drag reduction of civil transport wing fuselage configuration by using virtual game. The successful application validates efficiency of algorithms in solving complex aerodynamic optimization problem.

  11. Assessment of Voting Assistance Programs for Calendar Year 2011

    Science.gov (United States)

    2012-03-30

    TEXT 210 VOTER REGISTRATION PROGRAM Functional Area Manager: MRP -4 Point of Contact: MR. ROBERT WAGNER (DSN) 278-9511 (COML) (703) 784-5972 E-mail...2) 210 01 002 Did the MCVO subm[t a copy of his/her appointment letter to HQMC ( MRP -4)? Reference MCO 1742.1A W/CH 1-2, PAR 50 (1) 210 01 003...subordinate IVAO voting .assistance reports and submit one report to HQMC ( MRP -4) no later than 15 January of each year?’ Reference MCO 1742.1A W

  12. Basalt near-surface test facility test plans

    Energy Technology Data Exchange (ETDEWEB)

    Krug, A.D.

    1979-06-01

    The NSTF is under construction at Gable Mountain for in-situ testing, which will be conducted in two phases: Phase I, using electric heaters to simulate nuclear waste canisters in order to study the thermomechanical response of basalt; and Phase II, using spent fuel canisters. The tests planned for Phases I and II are described. (DLC)

  13. Observations during the first K West fuel shipping campaign

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-11-01

    Three fuel elements were shipped to the 300 Area hotcells during the first characterization shipping campaign from K West Basin. This document summarizes observations made during this campaign including the gas, liquid, and sludge content of the observed canisters. Included in an appendix is a detailed evaluation of fuel element condition for each canister opened

  14. System Specification for Immobilized High-Level Waste Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    CALMUS, R.B.

    2000-12-27

    This specification establishes the system-level functional, performance, design, interface, and test requirements for Phase 1 of the IHLW Interim Storage System, located at the Hanford Site in Washington State. The IHLW canisters will be produced at the Hanford Site by a Selected DOE contractor. Subsequent to storage the canisters will be shipped to a federal geologic repository.

  15. Transportation Costs as a Consideration in Air Force Contracts.

    Science.gov (United States)

    1979-03-10

    1978 Visits With Selected Government Contractors Bendix corporation South Bend , Indiana 14 Decemter 1978 Pratt & Whitney Aircraft and APPRO Wes t Palm ...Page 65 All 5~i.-355/NAVSt’l’ I ’Ul l - Ill (Kev .) - ’AFM 7i—2/ 15 March 1969 MCO l’-1600,1-IA/DSAIt 1500,3 fu’uuiii Ilis’ huilles i

  16. Including robustness in multi-criteria optimization for intensity-modulated proton therapy

    CERN Document Server

    Chen, Wei; Trofimov, Alexei; Madden, Thomas; Kooy, Hanne; Bortfeld, Thomas; Craft, David

    2011-01-01

    We present a method to include robustness into a multi-criteria optimization (MCO) framework for intensity-modulated proton therapy (IMPT). The approach allows one to simultaneously explore the trade-off between different objectives as well as the trade-off between robustness and nominal plan quality. In MCO, a database of plans each emphasizing different treatment planning objectives, is pre-computed to approximate the Pareto surface. An IMPT treatment plan that strikes the best balance between the different objectives can be selected by navigating on the Pareto surface. In our approach, robustness is integrated into MCO by adding robustified objectives and constraints to the MCO problem. Uncertainties of the robust problem are modeled by pre-calculated dose-influence matrices for a nominal scenario and a number of pre-defined error scenarios. A robustified objective represents the worst objective function value that can be realized for any of the error scenarios. The optimization method is based on a linear...

  17. 42 CFR 438.414 - Information about the grievance system to providers and subcontractors.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Information about the grievance system to providers..., DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL ASSISTANCE PROGRAMS MANAGED CARE Grievance System § 438.414 Information about the grievance system to providers and subcontractors. The MCO or PIHP...

  18. Guidelines For Integrating Population Education into Primary Education and Literacy Programmes.

    Science.gov (United States)

    Villanueva, Carmelita L., Ed.

    1989-01-01

    Presents guidelines for integrating population education (PE) into primary education and literacy programmes in the Asia and the Pacific region in the form of 12 steps. Recommends utilization and distribution methods for PE materials. Identifies issues and problems in the preparation and use of PE materials. (MCO)

  19. Population Education in the Nineties: A Quest for a Regional Programme Strategy in Asia and the Pacific.

    Science.gov (United States)

    Population Education in Asia and the Pacific Newsletter and Forum, 1991

    1991-01-01

    This study is a response to the emerging needs and requirements for population management with respect to regional strategies for population education. Addresses problems of illiteracy, poverty, and other impediments to population-growth management through the further training of those involved in the education process. (MCO)

  20. 42 CFR 438.208 - Coordination and continuity of care.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Coordination and continuity of care. 438.208... Improvement Access Standards § 438.208 Coordination and continuity of care. (a) Basic requirement—(1) General... MCO must meet the primary care coordination, identification, assessment, and treatment...

  1. Dactylogyrids (Monogenoidea: Polyonchoinea) parasitising the gills of snappers (Perciformes: Lutjanidae): species of Euryhaliotrema Kritsky & Boeger, 2002 from the golden snapper Lutjanus johnii (Bloch) off northern Australia, with a redescription of Euryhaliotrema johni (Tripathi, 1959) and descriptions of two new species.

    Science.gov (United States)

    Kritsky, Delane C; Diggles, Ben K

    2014-01-01

    Three species of Euryhaliotrema Kritsky & Boeger, 2002 (Monogenoidea: Dactylogyridae) were collected from the gills of four golden snapper Lutjanus johnii (Bloch) (Lutjanidae) from the marine and brackish waters off Darwin, Northern Territory, Australia. Type-specimens of Ancyrocephalus johni Tripathi, 1959 apparently have not survived and the possibility existed that the species was based on specimens representing more than one species. Euryhaliotrema johni (Tripathi, 1959) (sensu Young, 1968) was redescribed and determined to most likely represent A. johni, originally described from the River Hooghly, Diamond Harbour, India. Two new species were described. Euryhaliotrema longibaculoides n. sp. was most similar to Euryhaliotrema longibaculum (Zhukov, 1976) Kritsky & Boeger, 2002 from Lutjanus spp. from the western Atlantic Ocean. It differed from E. longibaculum by having a male copulatory organ (MCO) with an elongate comparatively delicate shaft and a bulbous base (MCO U- or J-shaped with funnel-shaped base in E. longibaculum). Based on the comparative morphology of the haptoral sclerites, Euryhaliotrema lisae n. sp. was most similar to Euryhaliotrema cryptophallus Kritsky & Yang, 2012 from the gills of the mangrove red snapper Lutjanus argentimaculatus (Forsskål) from the South China Sea. Euryhaliotrema lisae differed from E. cryptophallus by having a copulatory complex with an obvious weakly sclerotised J-shaped MCO (MCO cryptic, delicate, and with a shaft comprising about one counterclockwise ring in E. cryptophallus).

  2. 42 CFR 438.242 - Health information systems.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Health information systems. 438.242 Section 438.242... Measurement and Improvement Standards § 438.242 Health information systems. (a) General rule. The State must ensure, through its contracts, that each MCO and PIHP maintains a health information system that...

  3. 75 FR 6235 - Self-Regulatory Organizations; NASDAQ OMX BX, Inc.; Notice of Filing and Immediate Effectiveness...

    Science.gov (United States)

    2010-02-08

    ... Exchange Act Release No. 60886 (October 27, 2009), 74 FR 56897 (November 3, 2009) (SR-BX-2009-067). This.... 60950 (November 6, 2009), 74 FR 58666 (November 6, 2009) (SR-BX-2009-069). This proposal was effective.... AMAT Applied Materials Inc. MCO Moody's Corp. AMR AMR Corp. MET MetLife Inc. ANF Abercrombie & Fitch...

  4. 42 CFR 438.236 - Practice guidelines.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Practice guidelines. 438.236 Section 438.236 Public... Improvement Standards § 438.236 Practice guidelines. (a) Basic rule: The State must ensure, through its...) Adoption of practice guidelines. Each MCO and, when applicable, each PIHP and PAHP adopts...

  5. Major Combat Operations versus Stability Operations: Getting Army Priorities Correct

    Science.gov (United States)

    2011-05-19

    most likely encounter. Many defense analysts suggest that future conflict will be multimodal, combining various methods of warfare to increase both...fluid/maneuver http://usacac.leavenworth.amry.mil/BLOG/blogs/cac-t/archive/2009/04/09/semi-annual-ctc-conference-mco-and- comprehensive- aproach

  6. Surface characterization and direct bioelectrocatalysis of multicopper oxidases

    Energy Technology Data Exchange (ETDEWEB)

    Ivnitski, Dmitri M., E-mail: ivnitski@unm.ed [Chemical and Nuclear Engineering, University of New Mexico, Albuquerque 87131 (United States)] [Air Force Research Laboratory, AFRL/RXQL, Microbiology and Applied Biochemistry, Tyndall Air Force Base, FL 32403 (United States); Khripin, Constantine [Chemical and Nuclear Engineering, University of New Mexico, Albuquerque 87131 (United States); Luckarift, Heather R. [Air Force Research Laboratory, AFRL/RXQL, Microbiology and Applied Biochemistry, Tyndall Air Force Base, FL 32403 (United States)] [Universal Technology Corporation, 1270 N. Fairfield Road, Dayton, OH 45432 (United States); Johnson, Glenn R. [Air Force Research Laboratory, AFRL/RXQL, Microbiology and Applied Biochemistry, Tyndall Air Force Base, FL 32403 (United States); Atanassov, Plamen, E-mail: plamen@unm.ed [Chemical and Nuclear Engineering, University of New Mexico, Albuquerque 87131 (United States)

    2010-10-01

    Multicopper oxidases (MCO) have been extensively studied as oxygen reduction catalysts for cathodic reactions in biofuel cells. Theoretically, direct electron transfer between an enzyme and electrode offers optimal energy conversion efficiency providing that the enzyme/electrode interface can be engineered to establish efficient electrical communication. In this study, the direct bioelectrocatalysis of three MCO (Laccase from Trametes versicolor, bilirubin oxidase (BOD) from the fungi Myrothecium verrucaria and ascorbate oxidase (AOx) from Cucurbita sp.) was investigated and compared as oxygen reduction catalysts. Protein film voltammetry and electrochemical characterization of the MCO electrodes showed that DET had been successfully established in all cases. Atomic force microscopy imaging and force measurements indicated that enzyme was immobilized as a monolayer on the electrode surface. Evidence for three clearly separated anodic and cathodic redox events related to the Type 1 (T1) and the trinculear copper centers (T2, T3) of various MCO was observed. The redox potential of the T1 center was strongly modulated by physiological factors including pH, anaerobic and aerobic conditions and the presence of inhibitors.

  7. How Bright Is the Sun?

    Science.gov (United States)

    Berr, Stephen

    1991-01-01

    Presents a sequence of activities designed to allow eighth grade students to deal with one of the fundamental relationships that govern energy distribution. Activities guide students to measure light bulb brightness, discover the inverse square law, compare light bulb light to candle light, and measure sun brightness. (two references) (MCO)

  8. Production and dosimetric aspects of the potent Auger emitter Co-58m for targeted radionuclide therapy of small tumours

    DEFF Research Database (Denmark)

    Thisgaard, Helge; Elema, Dennis Ringkjøbing; Jensen, Mikael

    2011-01-01

    Based on theoretical calculations, the Auger emitter 58mCo has been identified as a potent nuclide for targeted radionuclide therapy of small tumors. During the production of this isotope, the coproduction of the long-lived ground state 58gCo is unfortunately unavoidable, as is ingrowth of the gr......Based on theoretical calculations, the Auger emitter 58mCo has been identified as a potent nuclide for targeted radionuclide therapy of small tumors. During the production of this isotope, the coproduction of the long-lived ground state 58gCo is unfortunately unavoidable, as is ingrowth...... of the ground state following the isomeric decay of 58mCo. The impact of 58gCo as a bþ- and c-emitting impurity should be included in the dosimetric analysis. The purpose of this study was to investigate this critical part of dosimetry based on experimentally determined production yields of 58mCo and 58g...

  9. Permaculture: Dreamworld or Breakthrough?

    Science.gov (United States)

    Johns, Maria

    1992-01-01

    Compares present agriculture practices to permaculture farming techniques, presents a historical perspective of permaculture and where these techniques are being successfully practiced around the world. Inserts (vignettes) enumerate the principles of permaculture and the background of Bill Mollison who conceptualized this farming practice. (MCO)

  10. Science Experimenter: Experimenting with a Geiger Counter.

    Science.gov (United States)

    Mims, Forrest M., III

    1992-01-01

    Describes the use of geiger counters for scientific investigations and experiments. Presents information about background radiation, its sources and detection. Describes how geiger counters work and other methods of radiation detection. Provides purchasing information for geiger counters, related computer software and equipment. (MCO)

  11. Finding Common Ground: Weed Management in Lincoln County, Montana.

    Science.gov (United States)

    Tonner, Carol

    1992-01-01

    Describes a personal experience in the effort to avoid widespread herbicide spraying. Provides insights for building a successful campaign: involvement, finding support, acceptance of differences of opinion, autonomy from political factions, and not assuming people are closed to healthier alternatives. (MCO)

  12. 42 CFR 438.808 - Exclusion of entities.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Exclusion of entities. 438.808 Section 438.808... Exclusion of entities. (a) General rule. FFP is available in payments under MCO contracts only if the State..., directly or indirectly, for the furnishing of health care, utilization review, medical social work,...

  13. Disposable Diapers Are OK.

    Science.gov (United States)

    Poore, Patricia

    1992-01-01

    A personal account of measuring the pros and cons of disposable diaper usage leads the author to differentiate between a garbage problem and environmental problem. Concludes the disposable diaper issue is a political and economic issue with a local environmental impact and well within our abilities to manage. (MCO)

  14. The Green Consumer Is Still Somewhat Pale.

    Science.gov (United States)

    Labovitz, David

    1993-01-01

    Calls for consumer initiated education concerning environmentally ethical buying decisions. Presents a historical depiction of American Consumerism associated with food consumption, marketing strategies, fads, the environmental movement, and American buying habits. Discusses government definitions and ensuing ambiguity of product terminology. (MCO)

  15. Design package for fuel retrieval system fuel handling tool modification

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1999-03-17

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

  16. Design package for fuel retrieval system fuel handling tool modification

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1998-11-09

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

  17. For Sale: Nuclear Waste Sites--Anyone Buying?

    Science.gov (United States)

    Hancock, Don

    1992-01-01

    Explores why the United States Nuclear Waste Program has been unable to find a volunteer state to host either a nuclear waste repository or monitored retrieval storage facility. Discusses the Department of Energy's plans for Nevada's Yucca Mountain as a repository and state and tribal responses to the plan. (21 references) (MCO)

  18. Biosphere 2: The True Story.

    Science.gov (United States)

    O'Keeffe, Michael

    1992-01-01

    Discusses the history and current developments of the Biosphere 2 Project, a prototype for enclosed self-sustaining structures for space colonization built in the Arizona Desert. Biosphere 2 was created to educate and provide solutions to environmental problems and revenue from research. (MCO)

  19. The Old Jalopy Races into the Future.

    Science.gov (United States)

    Considine, Tim

    1993-01-01

    Discusses alternative transportation technological advances in speed, range, battery strategies, and safety facilitated by motor car racing. Presents a historical perspective of the development of steam, electric and gas-powered vehicles and modern versions of electric, and mixed power source cars being tested today. (MCO)

  20. HARDMAN Comparability Analysis Methodology Guide. Volume 1. Manager’s Guide

    Science.gov (United States)

    1985-04-01

    34 ■ •wvrrvT-rmi iwi’jm’wwuj’ VA1 !TffTW.’ VJ HJHi H) Vf! qilw.M »’’’«S1* ^LWI ■l’unjuv11^ ’Jgf mff?w^P^^?^ gfff ^ V." 4 ^. Hy>, •"A MCO MEEI

  1. 支持多接口的NEMO实现与测试%IMPLEMENTATION AND TEST OF MULTIPLE INTERFACES-SUPPORTED NEMO

    Institute of Scientific and Technical Information of China (English)

    邱陆威; 高德云; 周华春

    2012-01-01

    Network Mobility ( NEMO) Basic Protocol is able to allow mobile nodes within sub-network to maintain continuity of the session when moving. Multiple care-of address (MCoA) protocol allows a mobile node to register multiple care-of addresses simultaneously. In this paper we integrate the protocols of NEMO and MCoA and study the implementation of the function of NEMO supported with multiple interfaces, and design the experiments to verify this function and analyse its performance test.%子网移动性(NEMO)基本支持协议可以让移动子网内部的节点在移动时依然保持会话的连续性,多转交地址协议(MCoA)允许一个移动节点同时注册多个转交地址.集成NEMO和MCoA协议,研究了支持多接口的NEMO功能的实现,并设计实验进行了功能验证和性能测试分析.

  2. 42 CFR 438.10 - Information requirements.

    Science.gov (United States)

    2010-10-01

    ... enrollee may obtain those benefits, any cost sharing, and how transportation is provided. For a counseling.... For a counseling or referral service that the MCO, PIHP, PAHP, or PCCM does not cover because of moral... service. (2) Advance directives, as set forth in § 438.6(i)(2). (3) Additional information that...

  3. 40 CFR 91.419 - Raw emission sampling calculations.

    Science.gov (United States)

    2010-07-01

    ... mass flow rate , MHCexh = Molecular weight of hydrocarbons in the exhaust; see the following equation: MHCexh = 12.01 + 1.008 × α Where: α=Hydrocarbon/carbon atomic ratio of the fuel. Mexh=Molecular weight of..., calculated from the following equation: ER04OC96.019 WCO = Mass rate of CO in exhaust, MCO = Molecular...

  4. [pi] Backbonding in Carbonyl Complexes and Carbon-Oxygen Stretching Frequencies: A Molecular Modeling Exercise

    Science.gov (United States)

    Montgomery, Craig D.

    2007-01-01

    An exercise is described that has illustrated the effect of various factors on [pi] backbonding to carbonyl ligands, where the students can view the molecular orbitals corresponding to the M-CO [pi] interaction as well as the competing interaction between the metal and co-ligands. The visual and hands-on nature of the modeling exercise has helped…

  5. Semantic programming model-based design

    NARCIS (Netherlands)

    Rovers, Kenneth C.; Kuper, Jan; Smit, Gerard J.M.

    2008-01-01

    For a generic flexible efficient array antenna receiver platform a hierarchical tiled architecture has been proposed, giving a heterogeneous multi-processor system-on-chip (MPSoC), multiple chips on a board (MCoB) and multiple boards in a system (MBiS). A wide range of MPSoCs are predicted to be use

  6. Bacterial biodegradation of melamine-contaminated aged soil: influence of different pre-culture media or addition of activation material.

    Science.gov (United States)

    Hatakeyama, Takashi; Takagi, Kazuhiro

    2016-08-01

    This study aimed to investigate the biodegrading potential of Arthrobacter sp. MCO, Arthrobacter sp. CSP, and Nocardioides sp. ATD6 in melamine-contaminated upland soil (melamine: approx. 10.5 mg/kg dry weight) after 30 days of incubation. The soil sample used in this study had undergone annual treatment of lime nitrogen, which included melamine; it was aged for more than 10 years in field. When R2A broth was used as the pre-culture medium, Arthrobacter sp. MCO could degrade 55 % of melamine after 30 days of incubation, but the other strains could hardly degrade melamine (approximately 25 %). The addition of trimethylglycine (betaine) in soil as an activation material enhanced the degradation rate of melamine by each strain; more than 50 % of melamine was degraded by all strains after 30 days of incubation. In particular, strain MCO could degrade 72 % of melamine. When the strains were pre-cultured in R2A broth containing melamine, the degradation rate of melamine in soil increased remarkably. The highest (72 %) melamine degradation rate was noted when strain MCO was used with betaine addition.

  7. Solid Waste Reduction--A Hands-on Study.

    Science.gov (United States)

    Wiessinger, Diane

    1991-01-01

    This lesson plan uses grocery shopping to demonstrate the importance of source reduction in the handling of solid waste problems. Students consider different priorities in shopping (convenience, packaging, and waste reduction) and draw conclusions about the relationship between packaging techniques and solid waste problems. (MCO)

  8. Conceptual design for PSP mounting bracket

    Energy Technology Data Exchange (ETDEWEB)

    Ransom, G.; Stein, R. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States)

    1991-12-31

    Protective structural packages (PSP`s or overpacks) used to ship 2 1/2-ton UF{sub 6} product cylinders are bolted to truck trailers. All bolts penetrate two longitudinal rows of wooden planks. Removal and replacement is required at various intervals for maintenance and routine testing. A conceptual design is presented for mounting brackets which would securely attach PSP`s to trailer frames, reduce removal and replacement time, and minimize risk of personnel injury.

  9. Measurement of pH of the Compacted Bentonite under the Reducing Condition

    OpenAIRE

    Nessa, Syeda Afsarun; Idemitsu, Kazuya; Yamasaki, Yosuke; Inagaki, Yaohiro; Arima, Tatsumi

    2007-01-01

    Compacted bentonite and carbon steel have been considered as the good buffer and over-pack materials in the repositories of high-level radioactive waste disposal. Sodium bentonite, Kunipia-F contains approximately 95wt% of montmorillonite. It has a high cation-exchange capacity and a high specific surface area, and its properties determine the behavior of bentonite. The pH of the pore water in compacted bentonite is an extremely important parameter because of its influence on radionuclide sol...

  10. Assessment of alternative disposal concepts

    Energy Technology Data Exchange (ETDEWEB)

    Autio, J.; Saanio, T.; Tolppanen, P. [Saanio and Riekkola Consulting Engineers, Helsinki (Finland); Raiko, H.; Vieno, T. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1996-12-01

    Four alternative repository designs for the disposal of spent nuclear in the Finnish crystalline bedrock were assessed in the study. The alternatives were: (1) the basic KBS-3 design in which copper canisters are emplaced in vertical deposition holes bored in the floors of horizontal tunnels, (2) the KBS-3-2C design with two canisters in a deposition hole, (3) Short Horizontal Holes (SHH) in the side walls of the tunnels, and (4) the Medium Long Holes (MLH) concept in which approximately 25 canisters are emplaced in a horizontal deposition hole about 200 metres in length bored between central and side tunnels. In all the alternatives considered, the thickness of the layer of compacted bentonite between copper canister and bedrock is 35 cm. Two different copper canister designs were also assessed. Technical feasibility and flexibility, post-closure safety and repository cost were assessed for each of the alternative canister and repository designs. On the basis of this assessment it is recommended that further development and studies should focus on the vacuum- or inert gas-filled cast insert type copper canister and the basic KBS-3 type repository design with a single canister in a vertical deposition hole. The KBS-3 design is robust and flexible and provides excellent post-closure safety. The transfer, emplacement and sealing operations are technically uncomplicated. The alternative options assessed do not offer any significant benefits in safety or cost over the basic design, but they are technically more complex and also in some respects more vulnerable to malfunction during the emplacement of canisters and buffer, as well as common mode failures. (60 refs.).

  11. Effects of brine migration on waste storage systems. Final report. [Thermomechanical effects

    Energy Technology Data Exchange (ETDEWEB)

    Gaffney, E.S.; Nickell, R.E.

    1979-05-15

    Processes which can lead to mobilization of brine adjacent to spent fuel or nuclear waste canisters and some of the thermomechanical consequences have been investigated. Velocities as high as 4 x 10/sup -7/ m s/sup -1/ (13 m y/sup -1/) are calculated at the salt/canister boundary. As much as 40 liters of pure NaCl brine could accumulate around each canister during a 10-year storage period. Accumulations of bittern brines would probably be less, in the range of 2 to 5 liters. With 0.5% water, NaCl brine accumulation over a 10-year storage cycle around a spent fuel canister producing 0.6 kW of heat is expected to be less than 1 liter for centimeter-size inclusions and less than 0.5 liter for millimeter-size inclusions. For bittern brines, about 25 years would be required to accumulate 0.4 liter. The most serious mechanical consequence of brine migration would be the increased mobility of the waste canister due to pressure solution. In pressure solution enhanced deformation, the existence of a thin film of fluid either between grains or between media (such as between a canister and the salt) provides a pathway by which the salt can be redistributed leading to a marked increase in strain rates in wet rock relative to dry rock. In salt, intergranular water will probably form discontinuous layers rather than films so that they would dominate pressure solution. A mathematical model of pressure solution indicates that pressure solution will not lead to appreciable canister motions except possibly in fine grained rocks (less than 10/sup -4/ m). In fine grained salts, details of the contact surface between the canister and the salt bed may lead to large pressure solution motions. A numerical model indicates that heat transfer in the brine layer surrounding a spent fuel canister is not conduction dominated but has a significant convective component.

  12. MO-G-304-04: Generating Well-Dispersed Representations of the Pareto Front for Multi-Criteria Optimization in Radiation Treatment Planning

    Energy Technology Data Exchange (ETDEWEB)

    Kirlik, G; Zhang, H [University of Maryland School of Medicine, Baltimore, MD (United States)

    2015-06-15

    Purpose: To present a novel multi-criteria optimization (MCO) solution approach that generates well-dispersed representation of the Pareto front for radiation treatment planning. Methods: Different algorithms have been proposed and implemented in commercial planning software to generate MCO plans for external-beam radiation therapy. These algorithms consider convex optimization problems. We propose a grid-based algorithm to generate well-dispersed treatment plans over Pareto front. Our method is able to handle nonconvexity in the problem to deal with dose-volume objectives/constraints, biological objectives, such as equivalent uniform dose (EUD), tumor control probability (TCP), normal tissue complication probability (NTCP), etc. In addition, our algorithm is able to provide single MCO plan when clinicians are targeting narrow bounds of objectives for patients. In this situation, usually none of the generated plans were within the bounds and a solution is difficult to identify via manual navigation. We use the subproblem formulation utilized in the grid-based algorithm to obtain a plan within the specified bounds. The subproblem aims to generate a solution that maps into the rectangle defined by the bounds. If such a solution does not exist, it generates the solution closest to the rectangle. We tested our method with 10 locally advanced head and neck cancer cases. Results: 8 objectives were used including 3 different objectives for primary target volume, high-risk and low-risk target volumes, and 5 objectives for each of the organs-at-risk (OARs) (two parotids, spinal cord, brain stem and oral cavity). Given tight bounds, uniform dose was achieved for all targets while as much as 26% improvement was achieved in OAR sparing comparing to clinical plans without MCO and previously proposed MCO method. Conclusion: Our method is able to obtain well-dispersed treatment plans to attain better approximation for convex and nonconvex Pareto fronts. Single treatment plan can

  13. 马来酸酐改性蓖麻油制备耐光性聚氨酯复鞣剂--乳液性能研究%Preparation of Sunproof Polyurethane Retanning Agents with Maleic Anhydride Modified Castor Oil--Study of the Emulsions Properties

    Institute of Scientific and Technical Information of China (English)

    鲍利红; 兰云军; 张淑芬

    2006-01-01

    用马来酸酐改性蓖麻油(MCO)合成了一系列不同组成的耐光性聚氨酯复鞣剂水乳液(MC-PUR),研究了复鞣剂中-COOH质量分数、n(NCO)/n(OH)、m(MCO)/m(PEG1000)对乳液电导率、黏度、临界聚沉值(CC.C)、耐酸稳定性的影响.结果表明,随-COOH质量分数从3%增大到7%,电导率从1 556 μs/cm增大到3 435 μs/cm,黏度先从168 mPa · s增大到224 mPa · s,后又降低到85 mPa · s,当w(-COOH)=5%时,黏度达到最大值;随n(NCO)/n(OH)从0.5增大到0.9,电导率从2 943 μs/cm降到2 464 μs/cm,黏度从428 m Pa · s降到224 mPa · s;随m(MCO): m(PEG1000)从1: 1增大到3: 0,黏度从224 mPa · s降到67 mPa · s;CC.C随-COOH质量分数增加和m(MCO)/m(PEG1000)增大而降低;耐酸稳定性结果证明,当w(-COOH)=3%~5%,m(MCO): m(PEG1000)=1: 1,n(NCO)/n(OH)=0.5~0.9时,MC-PUR适用于皮革复鞣工序.

  14. Effective combination of DIC, AE, and UPV nondestructive techniques on a scaled model of the Belgian nuclear waste container

    Science.gov (United States)

    Iliopoulos, Sokratis N.; Areias, Lou; Pyl, Lincy; Vantomme, John; Van Marcke, Philippe; Coppens, Erik; Aggelis, Dimitrios G.

    2015-03-01

    Protecting the environment and future generations against the potential hazards arising from high-level and heat emitting radioactive waste is a worldwide concern. Following this direction, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials has come up with the reference design which considers the geological disposal of the waste in purely indurated clay. In this design the wastes are first post-conditioned in massive concrete structures called Supercontainers before being transported to the underground repositories. The Supercontainers are cylindrical structures which consist of four engineering barriers that from the inner to the outer surface are namely: the overpack, the filler, the concrete buffer and possibly the envelope. The overpack, which is made of carbon steel, is the place where the vitrified wastes and spent fuel are stored. The buffer, which is made of concrete, creates a highly alkaline environment ensuring slow and uniform overpack corrosion as well as radiological shielding. In order to evaluate the feasibility to construct such Supercontainers two scaled models have so far been designed and tested. The first scaled model indicated crack formation on the surface of the concrete buffer but the absence of a crack detection and monitoring system precluded defining the exact time of crack initiation, as well as the origin, the penetration depth, the crack path and the propagation history. For this reason, the second scaled model test was performed to obtain further insight by answering to the aforementioned questions using the Digital Image Correlation, Acoustic Emission and Ultrasonic Pulse Velocity nondestructive testing techniques.

  15. Hydride heat pump with heat regenerator

    Science.gov (United States)

    Jones, Jack A. (Inventor)

    1991-01-01

    A regenerative hydride heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system. A series of at least four canisters containing a lower temperature performing hydride and a series of at least four canisters containing a higher temperature performing hydride is provided. Each canister contains a heat conductive passageway through which a heat transfer fluid is circulated so that sensible heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

  16. Evaluation of coupled thermo-hydro-mechanical phenomena in the near field for geological disposal of high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Chijimatsu, Masakazu; Fujita, Tomoo; Sugita, Yutaka; Taniguchi, Wataru [Japan Nuclear Cycle Development Inst., Tokai Works, Waste Management and Fuel Cycle Research Center, Waste Isolation Research Division, Barrier Performance Group, Tokai, Ibaraki (Japan)

    2000-01-01

    Geological disposal of high-level radioactive waste (HLW) in Japan is based on a multibarrier system composed of engineered and natural barriers. The engineered barriers are composed of vitrified waste confined within a canister, overpack and buffer material. Highly compacted bentonite clay is considered one of the most promising candidate buffer material mainly because of its low hydraulic conductivity and high adsorption capacity of radionuclides. In a repository of HLW, complex thermal, hydraulic and mechanical (T-H-M) phenomena will take place, involving the interactive processes between radioactive decay heat from the vitrified waste, infiltration of ground water and stress generation due to the earth pressure, the thermal loading and the swelling pressure of the buffer material. In order to evaluate the performance of the buffer material, the coupled T-H-M behaviors within the compacted bentonite have to be modelled. Before establishing a fully coupled T-H-M model, the mechanism of each single phenomenon or partially coupled phenomena should be identified. Furthermore, in order to evaluate the coupled T-H-M phenomena, the analysis model was developed physically and numerically and the adequacy and the applicability was tested though the engineered scale laboratory test and in-situ test. In this report, the investigative results for the development of coupled T-H-M model were described. This report consists of eight chapters. In Chapter 1, the necessity of coupled T-H-M model in the geological disposal project of the high-level radioactive waste was described . In Chapter 2, the laboratory test results of the rock sample and the buffer material for the coupled T-H-M analysis were shown. The rock samples were obtained from the in-situ experimental site at Kamaishi mine. As the buffer material, bentonite clay (Kunigel V1 and Kunigel OT-9607) and bentonite-sand mixture were used. In Chapter 3, in-situ tests to obtain the rock property were shown. As in-situ tests

  17. Evaluation of coupled thermo-hydro-mechanical phenomena in the near field for geological disposal of high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Chijimatsu, Masakazu; Fujita, Tomoo; Sugita, Yutaka; Taniguchi, Wataru [Japan Nuclear Cycle Development Inst., Tokai Works, Waste Management and Fuel Cycle Research Center, Waste Isolation Research Division, Barrier Performance Group, Tokai, Ibaraki (Japan)

    2000-01-01

    Geological disposal of high-level radioactive waste (HLW) in Japan is based on a multibarrier system composed of engineered and natural barriers. The engineered barriers are composed of vitrified waste confined within a canister, overpack and buffer material. Highly compacted bentonite clay is considered one of the most promising candidate buffer material mainly because of its low hydraulic conductivity and high adsorption capacity of radionuclides. In a repository of HLW, complex thermal, hydraulic and mechanical (T-H-M) phenomena will take place, involving the interactive processes between radioactive decay heat from the vitrified waste, infiltration of ground water and stress generation due to the earth pressure, the thermal loading and the swelling pressure of the buffer material. In order to evaluate the performance of the buffer material, the coupled T-H-M behaviors within the compacted bentonite have to be modelled. Before establishing a fully coupled T-H-M model, the mechanism of each single phenomenon or partially coupled phenomena should be identified. Furthermore, in order to evaluate the coupled T-H-M phenomena, the analysis model was developed physically and numerically and the adequacy and the applicability was tested though the engineered scale laboratory test and in-situ test. In this report, the investigative results for the development of coupled T-H-M model were described. This report consists of eight chapters. In Chapter 1, the necessity of coupled T-H-M model in the geological disposal project of the high-level radioactive waste was described . In Chapter 2, the laboratory test results of the rock sample and the buffer material for the coupled T-H-M analysis were shown. The rock samples were obtained from the in-situ experimental site at Kamaishi mine. As the buffer material, bentonite clay (Kunigel V1 and Kunigel OT-9607) and bentonite-sand mixture were used. In Chapter 3, in-situ tests to obtain the rock property were shown. As in-situ tests

  18. Novel Long-Term CO2 Removal System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Current Technology for CO2 removal from enclosed air of spacecraft utilizes LiOH canisters for CO2 absorption. This absorption is irreversible so longer flights...

  19. CO2 Removal from Mars EMU Project

    Data.gov (United States)

    National Aeronautics and Space Administration — CO2 control for during ExtraVehicular Activity (EVA) on mars is challenging. Lithium hydroxide (LiOH) canisters have impractical logistics penalties, and regenerable...

  20. 75 FR 12315 - Pacific Gas and Electric Company; Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2010-03-15

    ... MPC-32 canisters and, to allow linear interpolation for some enrichments consistent with the Holtec... conditions in the annular gap between the MPC and the transfer cask depending on which drying process is...

  1. Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste

    Institute of Scientific and Technical Information of China (English)

    H.G. Zhao; H. Shao; H. Kunz; J. Wang; R. Su; Y.M. Liu

    2014-01-01

    For deep geological disposal of high-level radioactive waste (HLW) in granite, the temperature on the HLW canisters is commonly designed to be lower than 100◦C. This criterion dictates the dimension of the repository. Based on the concept of HLW disposal in vertical boreholes, thermal process in the near field (host rock and buffer) surrounding HLW canisters has been simulated by using different methods. The results are drawn as follows:(a) the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperature field;(b) the thermal properties and variations of the host rock, the engineered buffer, and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation;(c) the gaps width and the filling by water or air determine the temperature offsets between them.

  2. Conceptual design criteria for facilities for geologic disposal of radioactive wastes in salt formations

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The facility design requirements and criteria discussed are: general codes, standards, specifications, and regulations; site criteria; land improvements criteria, low-level waste facility criteria; canistered waste facility criteria; support facilities criteria; and utilities and services criteria. (LK)

  3. Study of a Fractured Full Scale Inactive R7T7 Type Glass

    Institute of Scientific and Technical Information of China (English)

    GIN; Stephane

    2008-01-01

    <正>The nuclear glass block is fractured after it is poured into the canister. Aqueous alteration of glass involves essentially surface mechanisms; hence it is great importance to determine the surface area of the

  4. 42 CFR 84.1131 - Respirators; required components.

    Science.gov (United States)

    2010-10-01

    ... SAFETY AND HEALTH RESEARCH AND RELATED ACTIVITIES APPROVAL OF RESPIRATORY PROTECTIVE DEVICES Dust, Fume... noseclip, hood, or helmet; (2) Filter unit, canister with filter, or cartridge with filter; (3) Harness;...

  5. Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste

    Directory of Open Access Journals (Sweden)

    H.G. Zhao

    2014-02-01

    Full Text Available For deep geological disposal of high-level radioactive waste (HLW in granite, the temperature on the HLW canisters is commonly designed to be lower than 100 °C. This criterion dictates the dimension of the repository. Based on the concept of HLW disposal in vertical boreholes, thermal process in the near field (host rock and buffer surrounding HLW canisters has been simulated by using different methods. The results are drawn as follows: (a the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperature field; (b the thermal properties and variations of the host rock, the engineered buffer, and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation; (c the gaps width and the filling by water or air determine the temperature offsets between them.

  6. Environmental permits and approvals plan for high-level waste interim storage, Project W-464

    Energy Technology Data Exchange (ETDEWEB)

    Deffenbaugh, M.L.

    1998-05-28

    This report discusses the Permitting Plan regarding NEPA, SEPA, RCRA, and other regulatory standards and alternatives, for planning the environmental permitting of the Canister Storage Building, Project W-464.

  7. Reactive Rendezvous and Docking Sequencer Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Mars Sample Return poses some of the most challenging operational activities of any NASA deep space mission. Rendezvous of a vehicle with a sample canister in order...

  8. The Subject Headings of the Morris Swett Library, USAFAS. Revised.

    Science.gov (United States)

    1980-05-15

    Rifle. Ammunition, Beacon. See Iluminants - Beacon shells. AMMUNITION, BIOLOGICAL. AMMUNITION, BLANK. x Hlank armunition. AMMUNITION, CANISTER. (UL 400...illumination x Star shells - Beacon shells x Ammunition, Beacon - Candles - Flares - Flares, Aeroplanes x Aeroplanes - Flares - Iluminating shells (UL 440.1

  9. 50 CFR 32.28 - Florida.

    Science.gov (United States)

    2010-10-01

    ... report card and place it in an entrance fee canister each day prior to exiting the refuge. 12. All youth... after the Florida State Central Management Zone General Gun (antlered deer and wild hog) season...

  10. Research program to study the gamma radiation effects in Spanish bentonites; Programa de investigacion para estudiar los efectos de la radiacion gamma en bentonitas calcicas espanolas

    Energy Technology Data Exchange (ETDEWEB)

    Dies, J.; Tarrasa, F. [Universidad Politecnica de Catalunya (Spain); Cuevas de las, C.; Miralles, L.; Pueyo, J. J. [Universidad de Barcelona (Spain)

    2000-07-01

    The engineering barrier of a radioactive waste underground disposal facility, placed in a granitic host rock, will consist of a backfill of compacted bentonite blocks. At first, this material will be subjected to a gamma radiation field, from the waste canister, and heat from the spent fuel inside the canister. Moreover, any groundwater that reaches the repository will saturate the bentonite. For these reasons the performance of the engineered barrier must be carefully assessed in laboratory experiments. (Author)

  11. Dry spent fuel storage with the MACSTOR system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. [Atomic Energy of Canada Ltd., Montreal, PQ (Canada). CANDU Operations

    1996-10-01

    Atomic Energy of Canada Limited (AECL), and Transnuclear Inc. (TNI) began in 1989 the development of the concrete spent fuel storage system, called MACSTOR (Modular Air-Cooled Canister STORage) for use with LWR spent fuel assemblies. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The MACSTOR Module is a monolithic, shielded concrete vault structure that can accommodate up to 20 spent fuel canisters. Each canister typically holds up to 21 PWR or 44 BWR spent fuel assemblies with a nominal fuel burn up rate of 40,000 MWD/MTU and a 7 year minimum cooling period. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. Thus, the utility can be assured of both positive cooling of the fuel and verification of the integrity of the fuel confinement boundary. The structure is seismically designed and is capable of withstanding site design basis accident events. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. The MACSTOR system can economically address a wide range of storage capacity requirements. The modular concept allows for flexibility in determining each module`s capacity. Starting with 8 canisters, the capacity can be increased by increments of 4 up to 20 canisters. The MACSTOR system is also flexible in accommodating the various spent fuel types from such reactors as VVER-440, VVER-1000 and RBMK 1500. (J.P.N.)

  12. MACSTOR{trademark}: Dry spent fuel storage for the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Pattantyus, P. [AECL Candu, Montreal, Quebec (Canada); Hanson, A.S. [Transnuclear, Inc., Hawthorne, NY (United States)

    1993-12-31

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL`s successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection.

  13. General Purpose Satellites: a concept for affordable low earth orbit vehicles

    OpenAIRE

    Boyd, Austin W.; Fuhs, Allen E.

    1997-01-01

    A general purpose satellite has been designed which will be launched from the Space Shuttle using a NASA Get-Away-Special (GAS) canister. The design is based upon the use of a new extended GAS canister and a low profile launch mechanism. The satellite is cylindrical. measuring 19 inches in diameter and 35 inches long. The maximum vehicle weight is 250 pounds, of which 50 pounds is dedicated to user payloads. The remaining 200 pounds encompasses the satellite structure and support ...

  14. Treatment and final disposal of nuclear waste. Programme for encapsulation, deep geological disposal, and research, development and demonstration

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs.

  15. Treatment and disposal of radioactive wastes from nuclear power plants. Program for encapsulation, deep geologic deposition and research, development and demonstration; Kaernkraftavfallets behandling och slutfoervaring. Program foer inkapsling, geologisk djupfoervaring samt forskning, utveckling och demonstration

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs.

  16. Final report spent nuclear fuel retrieval system primary cleaning development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.

    1997-09-01

    Developmental testing of the primary cleaning station for spent nuclear fuel (SNF) and canisters is reported. A primary clean machine will be used to remove the gross sludge from canisters and fuel while maintaining water quality in the downstream process area. To facilitate SNF separation from canisters and minimize the impact to water quality, all canisters will be subjected to mechanical agitation and flushing with the Primary Clean Station. The Primary Clean Station consists of an outer containment box with an internally mounted, perforated wash basket. A single canister containing up to 14 fuel assemblies will be loaded into the wash basket, the confinement box lid closed, and the wash basket rotated for a fixed cycle time. During this cycle, basin water will be flushed through the wash basket and containment box to remove and entrain the sludge and carry it out of the box. Primary cleaning tests were performed to provide information concerning the removal of sludge from the fuel assemblies while in the basin canisters. The testing was also used to determine if additional fuel cleaning is required outside of the fuel canisters. Hydraulic performance and water demand requirements of the cleaning station were also evaluated. Thirty tests are reported in this document. Tests demonstrated that sludge can be dislodged and suspended sufficiently to remove it from the canister. Examination of fuel elements after cleaning suggested that more than 95% of the exposed fuel surfaces were cleaned so that no visual evidence of remained. As a result of testing, recommendations are made for the cleaning cycle. 3 refs., 16 figs., 4 tabs.

  17. Analysis of SKB MiniCan - Experiment 3

    Energy Technology Data Exchange (ETDEWEB)

    Smart, Nick; Rance, Andy; Reddy, Bharti; Fennell, Paul [AMEC (UK); Winsley, Robert [NDA (Country unknown/Code not available)

    2012-11-15

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB of Sweden is planning to use a system that consists of an outer copper canister and a cast iron insert (the KBS-3 concept). In 2007 Serco1 completed the set up of five model canister experiments at SKB's Aespoe laboratory and monitoring has continued since. The original aim of the model canister experiments was to examine how corrosion of the cast iron insert inside a copper canister would evolve with time, if water ingress through a small defect in the copper canister were to occur. Serco arranged manufacture and installation of five miniature copper canisters containing cast iron inserts, with 1 mm defects deliberately machined into the copper shell. The experiments use five small-scale model canisters (300 mm long x 150 mm diameter) that simulate the main features of the SKB canister design (hence the project name, 'MiniCan'). The main aim of the work is to examine how corrosion of the cast iron insert will evolve if a leak is present in the outer copper canister The experiments also included electrochemical equipment to monitor the corrosion behaviour of the model canisters in situ. In 2011 one of the experiments, Experiment 3, was removed f analysis. This report presents details of the procedures that were applied and the findings that were obtained from the analysis that was carried out on Experiment 3. To minimise exposure to air and to keep the contents of the experiment wet until the analysis was carried out, Experiment 3 was extracted from its borehole in August 2011 directly into a transfer tank that was filled with deaerated groundwater and placed in a purpose-built, water-filled and deoxygenated transfer flask. The transfer flask was then transported to the UK for dismantling and examination in a purpose-built anoxic glovebox that contained the appropriate lifting and cutting equipment for handling and sectioning the copper canister and the cast iron

  18. Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

    2013-07-15

    In this report, an analysis of the radiological consequences of potential accidents during disposal of spent nuclear fuel in deep boreholes is presented. The results presented should be seen as coarse estimates of possible radiological consequences of a canister being stuck in a borehole during disposal rather than being the results of a full safety analysis. In the concept for deep borehole disposal of spent nuclear fuel developed by Sandia National Laboratories, the fuel is assumed to be encapsulated in mild steel canisters and stacked between 3 and 5 km depth in boreholes that are cased with perforated mild steel casing tubes. The canisters are joined together by couplings to form strings of 40 canisters and lowered into the borehole. When a canister string has been emplaced in the borehole, a bridge plug is installed above the string and a 10 metres long concrete plug is cast on top of the bridge plug creating a floor for the disposal of the next sting. In total 10 canister strings, in all 400 canisters, are assumed to be disposed of at between 3 and 5 kilometres depth in one borehole. An analysis of potential accidents during the disposal operations shows that the potentially worst accident would be that a canister string is stuck above the disposal zone of a borehole and cannot be retrieved. In such a case, the borehole may have to be sealed in the best possible way and abandoned. The consequences of this could be that one or more leaking canisters are stuck in a borehole section with mobile groundwater. In the case of a leaking canister being stuck in a borehole section with mobile groundwater, the potential radiological consequences are likely to be dominated by the release of the so-called Instant Release Fraction (IRF) of the radionuclide inventory, i.e. the fraction of the radionuclides that as a consequence of the in-core conditions are present in the annulus between the fuel pellets and the cladding or on the grain boundaries of the UO{sub 2} matrix

  19. Transcripciones

    OpenAIRE

    2015-01-01

    Pieza n° 1: Pasacalle Pieza n° 2: Pasacalle Pieza n° 3: Juego Pieza n° 4: Tipac tipac Pieza n° 5: Cuatro esquinas Pieza n° 6: Alto ensayo – Waychaw Pieza n° 7: Alto ensayo - Siu sao Pieza n° 8: Pampa ensayo (versión 1) (Mco B y Mco 1 son ejecutadas al final de todas las versiones de Pampa ensayo) Pieza n° 9: Pampa ensayo (versión 2) Pieza n° 10: Pampa ensayo (versión 3) Pieza n° 11 : Pampa ensayo (versión 4) Pieza n° 12: Pampa ensayo (versión 5) Pieza n° 13: Pampa ensayo (versión ...

  20. Simultaneous determination of epinephrene and paracetamol at copper-cobalt oxide spinel decorated nanocrystalline zeolite modified electrodes.

    Science.gov (United States)

    Samanta, Subhajyoti; Srivastava, Rajendra

    2016-08-01

    In this study, CuCo2O4 and CuCo2O4 decorated nanocrystalline ZSM-5 materials were prepared. For comparative study, a series of MCo2O4 spinels were also prepared. Materials were characterized by the complementary combination of X-ray diffraction, N2-adsorption, UV-visible, and electron microscopic techniques. A simple and rapid method for the simultaneous determination of paracetamol and epinephrine at MCo2O4 spinels modified electrodes is presented in this manuscript. Among the materials investigated in this study, CuCo2O4 decorated nanocrystalline ZSM-5 exhibited the highest electrocatalytic activity with excellent stability, sensitivity, and selectivity. Analytical performance of the sensor was demonstrated in the determination of epinephrine and paracetamol in the commercial pharmaceutical samples.

  1. Backbone cyclization of a recombinant cystine-knot peptide by engineered Sortase A.

    Science.gov (United States)

    Stanger, Karen; Maurer, Till; Kaluarachchi, Harini; Coons, Mary; Franke, Yvonne; Hannoush, Rami N

    2014-11-28

    Cyclotides belong to the family of cyclic cystine-knot peptides and have shown promise as scaffolds for protein engineering and pharmacological modulation of cellular protein activity. Cyclotides are characterized by a cystine-knotted topology and a head-to-tail cyclic polypeptide backbone. While they are primarily produced in plants, cyclotides have also been obtained by chemical synthesis. However, there is still a need for methods to generate cyclotides in high yields to near homogeneity. Here, we report a biomimetic approach which utilizes an engineered version of the enzyme Sortase A to catalyze amide backbone cyclization of the recombinant cyclotide MCoTI-II, thereby allowing the efficient production of active homogenous species in high yields. Our results provide proof of concept for using engineered Sortase A to produce cyclic MCoTI-II and should be generally applicable to generating other cyclic cystine-knot peptides.

  2. An evaluation of the influence of primary care team functioning on the health of Medicare beneficiaries.

    Science.gov (United States)

    Roblin, Douglas W; Howard, David H; Junling Ren; Becker, Edmund R

    2011-04-01

    In service industries other than health care, unit employees who report a favorable service climate--characterized by commitment to a team concept and intrateam interactions that are supportive, collegial, and collaborative--have high levels of consumer satisfaction and work unit productivity. The authors evaluated whether similar primary care team (PCT) functioning influenced the short-term future health (SF-36) of elderly Medicare beneficiaries (N = 991) in a group model managed care organization (MCO). PCT functioning was assessed by surveys of practitioners and support staff on the MCO's 14 primary care practices and included measures of perceived task delegation, role collaboration, patient orientation, and team ownership. On average, patient physical and emotional health declined over 2 years. Medicare beneficiaries empanelled to relatively high functioning PCTs had significantly better physical and emotional health at 2 years following baseline assessment than those empanelled to relatively low functioning PCTs.

  3. Nuclear excitation functions of proton-induced reactions (Ep = 35-90 MeV) from Fe, Cu, and Al

    Science.gov (United States)

    Graves, Stephen A.; Ellison, Paul A.; Barnhart, Todd E.; Valdovinos, Hector F.; Birnbaum, Eva R.; Nortier, Francois M.; Nickles, Robert J.; Engle, Jonathan W.

    2016-11-01

    Fe, Cu, and Al stacked foils were irradiated by 90 MeV protons at the Los Alamos Neutron Science Center's Isotope Production Facility to measure nuclear cross sections for the production of medically relevant isotopes, such as 52gMn, 54Mn, 48Cr, 55Co, 58mCo and 57Ni. The decay of radioactive isotopes produced during irradiation was monitored using high-purity germanium gamma spectroscopy over the months following irradiation. Proton fluence was determined using the natAl(p,x)22Na, natCu(p,x)62Zn natCu(p,x)65Zn, and natCu(p,x)56Co monitor reactions. Calculated cross sections were compared against literature values and theoretical TALYS predictions. Notably this work includes the first reported independent cross section measurements of natCu(p,x)58mCo and natCu(p,x)58gCo.

  4. Fighting organized crime through open source intelligence: regulatory strategies of the CAPER Project

    OpenAIRE

    Casanovas, Pompeu; ARRAIZA, Juan; Melero, Felipe; González-Conejero, Jorge; Molcho, Gila; Cuadros, Montse

    2014-01-01

    OSINT stands for Open Source Intelligence. The CAPER project has built an OSINT solution oriented to the prevention of organised crime. We offer in this paper an overall view of some results, embedding into the system legal and ethical issues raised by the General Data Reform Package (GDRP) in Europe. We briefly describe CAPER architecture, workflow, functionalities, modules and ontologies (European LEAs Interoperability ELIO, and Multi-Lingual Crime Ontology MCO). This paper is focused on th...

  5. An overview on preseismic anomalies in LF radio signals revealed in Italy by wavelet analysis

    Directory of Open Access Journals (Sweden)

    A. Ermini

    2008-06-01

    Full Text Available Since 1996, the electric field strength of the two broadcasting stations MCO (f=216 kHz, southeast France and CZE (f=270 kHz, Czech Republic has been sampled every ten minutes by a receiver (AS located in central Italy. Here, we review the results obtained by a detailed analysis applied to the data recorded from February 1996 up to December 2004. At first, the daytime and nighttime data were extracted and then, in the daytime data, the data collected in winter were separated from those collected in summer. On the second step the wavelet transform was applied. The results of this analysis are radio anomalies detected as earthquake precursors both for MCO and CZE data. In particular, regarding the MCO data, the main result was the appearance of a very clear anomaly during May-August 1998, at daytime and at nighttime. Such an anomaly can be considered as a precursor of a seismic sequence started on August 15, 1998 with 17 earthquakes (M=2.2-4.6 on the Reatini mountains, a seismogenic zone located 30 km far from the AS receiver along the path MCO-AS. As concerns with the CZE data, the first result was obtained from the summer daytime data and it was the appearance of a very clear anomaly during August-September 1997, that can be considered a precursor of the two earthquakes with magnitude M=5.6 and M=5.9 that occurred on September 26 in the Umbria-Marche region (Central Italy. The second result was the appearance of an anomaly during February-March 1998, at daytime and at nighttime, that can be related to the preparatory phase of the strong (M=5.1-6.0 Slovenia seismic sequence that occurred in a zone lying in the middle of the CZE-AS path.

  6. 混凝土施工配合比的确定

    Institute of Scientific and Technical Information of China (English)

    李佳

    2011-01-01

    混凝土配合比设计计算:f=f+t×6,W/C=(а×f),(f+а×а×f,Mco=Mwo/(W/C),M+M+M+M=M,β=/(M+M)×100%,计算配合比为M:M:M:M.混凝土计算配合比的试配,混凝土配合比的调整与确定.

  7. Reactions and crystal structures of heterodinuclear complexes R3Sn-M(CO)5(M=Mn,Re)with some nitrogen ligands

    Institute of Scientific and Technical Information of China (English)

    Yong Qiang Ma; Ning Yin; Wen Jing Peng; Jing Li

    2009-01-01

    Absttact:Some reactions of R3SnMCO5M=Mn,Rewith CH3CN or pyridine were investigated to give complexes R3SnMnCO3LL'or R3SnMnCO4L by a facile mild method.X-ray diffractions analyses show that, in contrast to the phosphine ligand occupying in axial position, nitrogen ligands occupy equatorial position.

  8. Eccentricity pacing of eastern equatorial Pacific carbonate dissolution cycles during the Miocene Climatic Optimum

    Science.gov (United States)

    Kochhann, Karlos G. D.; Holbourn, Ann; Kuhnt, Wolfgang; Channell, James E. T.; Lyle, Mitch; Shackford, Julia K.; Wilkens, Roy H.; Andersen, Nils

    2016-09-01

    The Miocene Climatic Optimum (MCO; ~16.9 to 14.7 Ma) provides an outstanding opportunity to investigate climate-carbon cycle dynamics during a geologically recent interval of global warmth. We present benthic stable oxygen (δ18O) and carbon (δ13C) isotope records (5-12 kyr time resolution) spanning the late early to middle Miocene interval (18 to 13 Ma) at Integrated Ocean Drilling Program (IODP) Site U1335 (eastern equatorial Pacific Ocean). The U1335 stable isotope series track the onset and development of the MCO as well as the transitional climatic phase culminating with global cooling and expansion of the East Antarctic Ice Sheet at ~13.8 Ma. We integrate these new data with published stable isotope, geomagnetic polarity, and X-ray fluorescence (XRF) scanner-derived carbonate records from IODP Sites U1335, U1336, U1337, and U1338 on a consistent, astronomically tuned timescale. Benthic isotope and XRF scanner-derived CaCO3 records depict prominent 100 kyr variability with 400 kyr cyclicity additionally imprinted on δ13C and CaCO3 records, pointing to a tight coupling between the marine carbon cycle and climate variations. Our intersite comparison further indicates that the lysocline behaved in highly dynamic manner throughout the MCO, with >75% carbonate loss occurring at paleodepths ranging from ~3.4 to ~4 km in the eastern equatorial Pacific Ocean. Carbonate dissolution maxima coincide with warm phases (δ18O minima) and δ13C decreases, implying that climate-carbon cycle feedbacks fundamentally differed from the late Pleistocene glacial-interglacial pattern, where dissolution maxima correspond to δ13C maxima and δ18O minima. Carbonate dissolution cycles during the MCO were, thus, more similar to Paleogene hyperthermal patterns.

  9. Feasibility of constant dose rate VMAT in the treatment of nasopharyngeal cancer patients

    OpenAIRE

    Yu, Wenliang; Shang, Haijiao; Xie, Congying; Han, CE; Yi, Jinling; Zhou, Yongqiang; Jin, Xiance

    2014-01-01

    Purpose To investigate the feasibility of constant dose rate volumetric modulated arc therapy (CDR-VMAT) in the treatment of nasopharyngeal cancer (NPC) patients and to introduce rotational arc radiotherapy for linacs incapable of dose rate variation. Materials and methods Twelve NPC patients with various stages treated previously using variable dose rate (VDR) VMAT were enrolled in this study. CDR-VMAT, VDR-VMAT and mutlicriteria optimization (MCO) VMAT plans were generated for each patient ...

  10. [Effects of antiseptic on the analysis of greenhouse gases concentrations in lake water].

    Science.gov (United States)

    Xiao, Qi-Tao; Hu, Zheng-Hu; James, Deng; Xiao, Wei; Liu, Shou-Dong; Li, Xu-Hui

    2014-01-01

    To gain insight into antiseptic effects on the concentrations of CO2, CH4, and N2O in lake water, antisepetic (CuSO4 and HgCl2) were added into water sample, and concentrations of greenhouse gases were measured by the gas chromatography based on water equilibrium method. Experiments were conducted as following: the control group without antisepetic (CK), the treatment group with 1 mL CuSO4 solution (T1), the treatment group with 5 mL CuSO4 solution (T2), and the treatment group with 0.5 mL HgCl2 solution (T3). All groups were divided into two batches: immediately analysis (I), and after 2 days analysis (II). Results showed that CuSO4 and HgCl2 significantly increased CO2 concentration, the mean CO2 concentration (Mco2) of CK (I) and CK (II) were (11.5 +/- 1.47) micromol x L(-1) and (14.38 +/- 1.59) micromol x L(-1), respectively; the Mco2 of T1 (I) and T1 (II) were (376 +/- 70) micromol x L(-1) and (448 +/- 246.83) micromol x L(-1), respectively; the Mco2 of T2 (I) and T2 (II) were (885 +/- 51.53) micromol x L(-1) and (988.83 +/- 101.96) micromol x L(-1), respectively; the Mco2 of T3 (I) and T3 (II) were (287.19 +/- 30.01) micromol x L(-1) and (331.33 +/- 22.06) micromol x L(-1), respectively. The results also showed that there was no difference in CH4 and N2O concentrations among treatments. Water samples should be analyzed as soon as possible after pretreatment. Our findings suggest that adding antiseptic may lead an increase in CO2 concentration.

  11. Analysis of Promotion Rates to Lieutenant Colonel and Selection for Command for USMC Aviation Supply and Maintenance Officers

    Science.gov (United States)

    2011-12-01

    Commissioned Officer Accession Career MCO Marine Corps Order MCMAP Marine Corps Martial Arts Program MMCO Maintenance Material Control Officer...Marine Corps Martial Arts 16 Program (MCMAP), as well as any other program mandated by the OPSO. An AVNSUPO serving in the S-3 falls under the HQ...in any statistical analysis. Similar problems have prevented the USMC Martial Arts Center of Excellence from providing accurate numbers of trained

  12. SR 97. Alternative models project. Stochastic continuum modelling of Aberg

    Energy Technology Data Exchange (ETDEWEB)

    Widen, H. [Kemakta AB, Stockholm (Sweden); Walker, D. [INTERA KB/DE and S (Sweden)

    1999-08-01

    As part of studies into the siting of a deep repository for nuclear waste, Swedish Nuclear Fuel and Waste Management Company (SKB) has commissioned the Alternative Models Project (AMP). The AMP is a comparison of three alternative modelling approaches to bedrock performance assessment for a single hypothetical repository, arbitrarily named Aberg. The Aberg repository will adopt input parameters from the Aespoe Hard Rock Laboratory in southern Sweden. The models are restricted to an explicit domain, boundary conditions and canister location to facilitate the comparison. The boundary conditions are based on the regional groundwater model provided in digital format. This study is the application of HYDRASTAR, a stochastic continuum groundwater flow and transport-modelling program. The study uses 34 realisations of 945 canister locations in the hypothetical repository to evaluate the uncertainty of the advective travel time, canister flux (Darcy velocity at a canister) and F-ratio. Several comparisons of variability are constructed between individual canister locations and individual realisations. For the ensemble of all realisations with all canister locations, the study found a median travel time of 27 years, a median canister flux of 7.1 x 10{sup -4} m/yr and a median F-ratio of 3.3 x 10{sup 5} yr/m. The overall pattern of regional flow is preserved in the site-scale model, as is reflected in flow paths and exit locations. The site-scale model slightly over-predicts the boundary fluxes from the single realisation of the regional model. The explicitly prescribed domain was seen to be slightly restrictive, with 6% of the stream tubes failing to exit the upper surface of the model. Sensitivity analysis and calibration are suggested as possible extensions of the modelling study.

  13. SR 97 - Alternative models project. Discrete fracture network modelling for performance assessment of Aberg

    Energy Technology Data Exchange (ETDEWEB)

    Dershowitz, B.; Eiben, T. [Golder Associates Inc., Seattle (United States); Follin, S.; Andersson, Johan [Golder Grundteknik KB, Stockholm (Sweden)

    1999-08-01

    As part of studies into the siting of a deep repository for nuclear waste, Swedish Nuclear Fuel and Waste Management Company (SKB) has commissioned the Alternative Models Project (AMP). The AMP is a comparison of three alternative modeling approaches for geosphere performance assessment for a single hypothetical site. The hypothetical site, arbitrarily named Aberg is based on parameters from the Aespoe Hard Rock Laboratory in southern Sweden. The Aberg model domain, boundary conditions and canister locations are defined as a common reference case to facilitate comparisons between approaches. This report presents the results of a discrete fracture pathways analysis of the Aberg site, within the context of the SR 97 performance assessment exercise. The Aberg discrete fracture network (DFN) site model is based on consensus Aberg parameters related to the Aespoe HRL site. Discrete fracture pathways are identified from canister locations in a prototype repository design to the surface of the island or to the sea bottom. The discrete fracture pathways analysis presented in this report is used to provide the following parameters for SKB's performance assessment transport codes FARF31 and COMP23: * F-factor: Flow wetted surface normalized with regards to flow rate (yields an appreciation of the contact area available for diffusion and sorption processes) [TL{sup -1}]. * Travel Time: Advective transport time from a canister location to the environmental discharge [T]. * Canister Flux: Darcy flux (flow rate per unit area) past a representative canister location [LT{sup -1}]. In addition to the above, the discrete fracture pathways analysis in this report also provides information about: additional pathway parameters such as pathway length, pathway width, transport aperture, reactive surface area and transmissivity, percentage of canister locations with pathways to the surface discharge, spatial pattern of pathways and pathway discharges, visualization of pathways, and

  14. A synaptic input portal for a mapped clock oscillator model of neuronal electrical rhythmic activity

    Science.gov (United States)

    Zariffa, José; Ebden, Mark; Bardakjian, Berj L.

    2004-09-01

    Neuronal electrical oscillations play a central role in a variety of situations, such as epilepsy and learning. The mapped clock oscillator (MCO) model is a general model of transmembrane voltage oscillations in excitable cells. In order to be able to investigate the behaviour of neuronal oscillator populations, we present a neuronal version of the model. The neuronal MCO includes an extra input portal, the synaptic portal, which can reflect the biological relationships in a chemical synapse between the frequency of the presynaptic action potentials and the postsynaptic resting level, which in turn affects the frequency of the postsynaptic potentials. We propose that the synaptic input-output relationship must include a power function in order to be able to reproduce physiological behaviour such as resting level saturation. One linear and two power functions (Butterworth and sigmoidal) are investigated, using the case of an inhibitory synapse. The linear relation was not able to produce physiologically plausible behaviour, whereas both the power function examples were appropriate. The resulting neuronal MCO model can be tailored to a variety of neuronal cell types, and can be used to investigate complex population behaviour, such as the influence of network topology and stochastic resonance.

  15. Antagonist muscle co-activation of limbs in human infant crawling: A pilot study.

    Science.gov (United States)

    Xiong, Qi L; Wu, Xiao Y; Xiao, Nong; Zeng, Si Y; Wan, Xiao P; Zheng, Xiao L; Hou, Wen S

    2015-01-01

    Muscle Co-activation (MCo) is the simultaneous muscular activation of agonist and antagonist muscle groups, which provides adequate joint stability, movement accuracy during movement. Infant crawling is an important stage of motor function development that manifests non-synchronization growth and development of upper and lower limbs due to the well-known gross motor development principle of head to toe. However, the effect of MCo level for agonist and antagonist muscle groups on motor function development of limbs has not been previously reported. In this paper, sEMG signals were collected from triceps brachii (TB) and biceps brachii (BB), quadriceps femoris (QF) and hamstrings (HS) of limbs when fourteen infants were crawling at their self-selected speed. Antagonist muscle co-activation was evaluated by measuring two common indexes (co-activation index and Pearson's correlation coefficient).A significant difference was observed between upper limbs and lower limbs, but the relationship between MCo and speed of crawling was poor. This study is an opening for further investigation including a longitudinal study and compare against infant with CNS disorders.

  16. Bilirubin oxidase-like proteins from Podospora anserina: promising thermostable enzymes for application in transformation of plant biomass.

    Science.gov (United States)

    Xie, Ning; Ruprich-Robert, Gwenaël; Silar, Philippe; Chapeland-Leclerc, Florence

    2015-03-01

    Plant biomass degradation by fungi is a critical step for production of biofuels, and laccases are common ligninolytic enzymes envisioned for ligninolysis. Bilirubin oxidases (BODs)-like are related to laccases, but their roles during lignocellulose degradation have not yet been fully investigated. The two BODs of the ascomycete fungus Podospora anserina were characterized by targeted gene deletions. Enzymatic assay revealed that the bod1(Δ) and bod2(Δ) mutants lost partly a thermostable laccase activity. A triple mutant inactivated for bod1, bod2 and mco, a previously investigated multicopper oxidase gene distantly related to laccases, had no thermostable laccase activity. The pattern of fruiting body production in the bod1(Δ) bod2(Δ) double mutant was changed. The bod1(Δ) and bod2(Δ) mutants were reduced in their ability to grow on ligneous and cellulosic materials. Furthermore, bod1(Δ) and bod2(Δ) mutants were defective towards resistance to phenolic substrates and H2 O2 , which may also impact lignocellulose breakdown. Double and triple mutants were more affected than single mutants, evidencing redundancy of function among BODs and mco. Overall, the data show that bod1, bod2 and mco code for non-canonical thermostable laccases that participate in the degradation of lignocellulose. Thanks to their thermal stability, these enzymes may be more promising candidate for biotechnological application than canonical laccases.

  17. Effects of Degree of Surgical Correction for Flatfoot Deformity in Patient-Specific Computational Models.

    Science.gov (United States)

    Spratley, E M; Matheis, E A; Hayes, C W; Adelaar, R S; Wayne, J S

    2015-08-01

    A cohort of adult acquired flatfoot deformity rigid-body models was developed to investigate the effects of isolated tendon transfer with successive levels of medializing calcaneal osteotomy (MCO). Following IRB approval, six diagnosed flatfoot sufferers were subjected to magnetic resonance imaging (MRI) and their scans used to derive patient-specific models. Single-leg stance was modeled, constrained solely through physiologic joint contact, passive soft-tissue tension, extrinsic muscle force, body weight, and without assumptions of idealized mechanical joints. Surgical effect was quantified using simulated mediolateral (ML) and anteroposterior (AP) X-rays, pedobarography, soft-tissue strains, and joint contact force. Radiographic changes varied across states with the largest average improvements for the tendon transfer (TT) + 10 mm MCO state evidenced through ML and AP talo-1st metatarsal angles. Interestingly, 12 of 14 measures showed increased deformity following TT-only, though all increases disappeared with inclusion of MCO. Plantar force distributions showed medial forefoot offloading concomitant with increases laterally such that the most corrected state had 9.0% greater lateral load. Predicted alterations in spring, deltoid, and plantar fascia soft-tissue strain agreed with prior cadaveric and computational works suggesting decreased strain medially with successive surgical repair. Finally, joint contact force demonstrated consistent medial offloading concomitant with variable increases laterally. Rigid-body modeling thus offers novel advantages for the investigation of foot/ankle biomechanics not easily measured in vivo.

  18. Bio-based polyurethane prepared from Kraft lignin and modified castor oil

    Directory of Open Access Journals (Sweden)

    L. B. Tavares

    2016-11-01

    Full Text Available Current challenges highlight the need for polymer research using renewable natural sources as a substitute for petroleum-based polymers. The use of polyols obtained from renewable sources combined with the reuse of industrial residues such as lignin is an important agent in this process. Different compositions of polyurethane-type materials were prepared by combining technical Kraft lignin (TKL with castor oil (CO or modified castor oil (MCO1 and MCO2 to increase their reactivity towards diphenylmethane diisocyanate (MDI. The results indicate that lignin increases the glass transition temperature, the crosslinking density and improves the ultimate stress especially for those prepared from MCO2 and 30% lignin content from 8.2 MPa (lignin free to 23.5 MPa. Scanning electron microscopy (SEM micrographs of rupture surface after uniaxial tensile tests show ductile-to-brittle transition. The results show the possibility to develop polyurethane-type materials, varying technical grade Kraft lignin content, which cover a wide range of mechanical properties (from large elastic/low Young modulus to brittle/high Young modulus polyurethanes.

  19. Information Gathering Document 0321-1437-30-R-OG

    Energy Technology Data Exchange (ETDEWEB)

    Hollister, R

    2009-07-15

    Fines and turnings from machining depleted uranium (Dep-U), natural uranium (Nat-U), and Thorium-232, and stainless steel and aluminum. This IGO allows only small, oxidizable pieces of Dep-U/Nat-U/Th-232, with regulated metal contaminants below regulatory limits. Fines and turnings will be in 30 gallon vented drums immersed in mineral oil. The 30 gallon drums will be overpacked in 55 gallon vented drums. The waste will be stored on site until sent for stabilization & disposal with approved TSOFs.

  20. CANE FIBERBOARD DEGRADATION WITHIN THE 9975 SHIPPING PACKAGE DURING LONG-TERM STORAGE APPLICATION

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.; Dunn, K.; Hackney, B.

    2013-06-19

    The 9975 shipping package is used as part of the configuration for long-term storage of special nuclear materials in the K Area Complex at the Savannah River Site. The cane fiberboard overpack in the 9975 package provides thermal insulation, impact absorption and criticality control functions relevant to this application. The Savannah River National Laboratory has conducted physical, mechanical and thermal tests on aged fiberboard samples to identify degradation rates and support the development of aging models and service life predictions in a storage environment. This paper reviews the data generated to date, and preliminary models describing degradation rates of cane fiberboard in elevated temperature – elevated humidity environments.

  1. CH Packaging Operations for High Wattage Waste at LANL

    Energy Technology Data Exchange (ETDEWEB)

    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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. CH Packaging Operations for High Wattage Waste at LANL

    Energy Technology Data Exchange (ETDEWEB)

    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. CH Packaging Operations for High Wattage Waste at LANL

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. CH Packaging Operations for High Wattage Waste at LANL

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. 327 to 324 Pin tube shipment quality management process plan

    Energy Technology Data Exchange (ETDEWEB)

    HAM, J.E.

    1998-11-05

    The B and W Hanford Company's (BWHC) 327 Facility, in the 300 Area of the Hanford Site, is preparing to ship five Pin Tubes to the 324 Facility for storage and eventual disposition. The Pin Tubes consist of legacy fuel pin pieces and drillings. They will be over-packed in new Pin Tubes and transported to 324 in three shipments. Once received at 324, two of the shipments will be combined for storage as a fissionable material batch, and the other shipment will be added to an existing batch.

  7. Technical note: Can the sulfur hexafluoride tracer gas technique be used to accurately measure enteric methane production from ruminally cannulated cattle?

    Science.gov (United States)

    Beauchemin, K A; Coates, T; Farr, B; McGinn, S M

    2012-08-01

    An experiment was conducted to determine whether using ruminally cannulated cattle affects the estimate of enteric methane (CH(4)) emissions when using the sulfur hexafluoride (SF(6)) tracer technique with samples taken from a head canister. Eleven beef cattle were surgically fitted with several types of ruminal cannula (2C, 3C, 3C+washer, 9C; Bar Diamond, Parma, ID). The 2C and 3C models (outer and inner flanges with opposite curvature) had medium to high leakage, whereas the 9C models (outer and inner flanges with the same curvature) provided minimum to moderate leakage of gas. A total of 48 cow-day measurements were conducted. For each animal, a permeation tube containing sulfur hexafluoride (SF(6)) was placed in the rumen, and a sample of air from around the nose and mouth was drawn through tubing into an evacuated canister (head canister). A second sample of air was collected from outside the rumen near the cannula into another canister (cannula canister). Background concentrations were also monitored. The methane (CH(4)) emission was estimated from the daily CH(4) and SF(6) concentrations in the head canister (uncorrected). The permeation SF(6) release rate was then partitioned based on the proportion of the SF(6) concentration measured in the head vs. the cannula canister. The CH(4) emissions at each site were calculated using the two release rates and the two CH(4):SF(6) concentration ratios. The head and cannula emissions were summed to obtain the total emission (corrected). The difference (corrected - uncorrected) in CH4 emission was attributed to the differences in CH(4):SF(6) ratio at the 2 exit locations. The proportions of CH(4) and SF(6) recovered at the head were greater (P 0.05; 2C, 6% and 4%; 3C, 17% and 15%; 3C+washer, 19% and 14%). Uncorrected CH(4) emissions were ± 10% of corrected emissions for 53% of the cow-day measurements. Only when more than 80% of the SF(6) escaped via the rumen did the difference between the uncorrected and corrected

  8. Design of spent-fuel concrete pit dry storage and handling system

    Energy Technology Data Exchange (ETDEWEB)

    Tamaki, H.; Natsume, T.; Maruoka, K.; Yokoyama, T. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan)

    1998-07-01

    An advanced dry storage system design with highly improved storage efficiency of spent nuclear fuel has been developed. The new concept 'Concrete Pit Dry Storage System' realizes a safe and economical solution to an increasing demand of storing spent fuel assemblies (SFAs) generated from commercial nuclear power reactors. The system is basically composed of a large mass concrete module which has densely arranged pit boreholes, sealed canisters containing spent fuel assemblies and a canister handling system. The system is characterized by the following advantages compared with the existing concrete module type storage systems: higher storage efficiency can be achieved by the storage module filled with concrete which also gives a high shielding performance; simple handling technology is used for transfer and installation of the canisters at the storage facility as well as the transport cask of the canisters, surface contamination of the canister is prevented; lower radiation around the storage area is provided to reduce radiation exposure during handling and storage; high structural integrity of the facility is maintained by the concrete module with a simple construction ; the ventilation gallery introducing cooling air air to the bit borehole has an enough draft height to improve cooling performance of the system; a result of the design concept, the storage system can store higher burn-up SFAs with a short cooling period. (authors)

  9. Efficacy of backfilling and other engineered barriers in a radioactive waste repository in salt

    Energy Technology Data Exchange (ETDEWEB)

    Claiborne, H.C.

    1982-09-01

    In the United States, investigation of potential host geologic formations was expanded in 1975 to include hard rocks. Potential groundwater intrusion is leading to very conservative and expensive waste package designs. Recent studies have concluded that incentives for engineered barriers and 1000-year canisters probably do not exist for reasonable breach scenarios. The assumption that multibarriers will significantly increase the safety margin is also questioned. Use of a bentonite backfill for surrounding a canister of exotic materials was developed in Sweden and is being considered in the US. The expectation that bentonite will remain essentially unchanged for hundreds of years for US repository designs may be unrealistic. In addition, thick bentonite backfills will increase the canister surface temperature and add much more water around the canister. The use of desiccant materials, such as CaO or MgO, for backfilling seems to be a better method of protecting the canister. An argument can also be made for not using backfill material in salt repositories since the 30-cm-thick space will provide for hole closure for many years and will promote heat transfer via natural convection. It is concluded that expensive safety systems are being considered for repository designs that do not necessarily increase the safety margin. It is recommended that the safety systems for waste repositories in different geologic media be addressed individually and that cost-benefit analyses be performed.

  10. Thermal and mechanical analysis for the detailed model using submodel

    Energy Technology Data Exchange (ETDEWEB)

    Kuh, Jung Eui; Kang, Chul Hyung; Park, Jeong Hwa [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    A very big model is required for the TM analysis for HLRW repository, and also very small size of mesh is needed to simulate precisely main parts of analysis, e.g., canister, buffer, etc. However, it is practically impossible due to high memory size and computing time. In this report, a submodel concept in ABAQUS is used to handle this difficulty. A submodel concept is the part interested only is performed detailed modelling and this result is used as a boundary condition of full scale model. To follow this kind of computation procedure temperature distribution in buffer and canister could be computed precisely. This approach can be applied to TM analysis of buffer and canister, or a finite size of repository. 12 refs., 28 figs., 9 tabs. (Author)

  11. CASTOR GSF packaging design criteria

    Energy Technology Data Exchange (ETDEWEB)

    Burnside, M.E.

    1996-08-06

    Encapsulated vitrified materials (Isotopic Heat Sources) are currently stored in the Pacific Northwest National Laboratories (PNNL) 324 Building located in the 300 Area. As part of the 324 Building transition program, the vitrified material, encapsulated in stainless steel canisters, must be removed. These canisters were originally intended to be used by the German government, but are no longer desired. As part of an agreement with the German government, the Germans are providing the U.S. Department of Energy (DOE) with six (6) CASTOR GSF and four (4) GNS-12 casks.The canisters will be transported onsite in CASTOR GSF and GNS-12 casks for interim storage until final disposition of the material is determined.

  12. Radiation Heat Transfer Modeling Improved for Phase-Change, Thermal Energy Storage Systems

    Science.gov (United States)

    Kerslake, Thomas W.; Jacqmin, David A.

    1998-01-01

    Spacecraft solar dynamic power systems typically use high-temperature phase-change materials to efficiently store thermal energy for heat engine operation in orbital eclipse periods. Lithium fluoride salts are particularly well suited for this application because of their high heat of fusion, long-term stability, and appropriate melting point. Considerable attention has been focused on the development of thermal energy storage (TES) canisters that employ either pure lithium fluoride (LiF), with a melting point of 1121 K, or eutectic composition lithium-fluoride/calcium-difluoride (LiF-20CaF2), with a 1040 K melting point, as the phase-change material. Primary goals of TES canister development include maximizing the phase-change material melt fraction, minimizing the canister mass per unit of energy storage, and maximizing the phase-change material thermal charge/discharge rates within the limits posed by the container structure.

  13. Numerical simulation of temperature field in deep penetration laser welding of 5A06 aluminum cylinder

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Deep penetration laser welding temperature field of 5A06 aluminum alloy canister structure was simulated using the surface-body combination heat source model by ANSYS, which was made up of Gauss surface heat source model and Gauss revolved body heat source model. Convection, radiation and conduction were all considered during the simulation process. The thermal cycle curves of the points both on the shell outer surface and in the seam thickness direction were calculated. Simulated results agreed well with the experiment results. It concluded that the surface-body combination heat source model was fit for the temperature field simulation of deep penetration laser welding of the aluminum alloy canister structure. This method was proved to be an efficient way to predict the shape and dimension of welded joint for deep penetration laser welding of the aluminum alloy canister structure.

  14. TMI Fuel Characteristics for Disposal Criticality Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Larry L. Taylor

    2003-09-01

    This report documents the reported contents of the Three Mile Island Unit 2 (TMI-2) canisters. proposed packaging, and degradation scenarios expected in the repository. Most fuels within the U.S. Department of Energy spent nuclear fuel inventory deal with highly enriched uranium, that in most cases require some form of neutronic poisoning inside the fuel canister. The TMI-2 fuel represents a departure from these fuel forms due to its lower enrichment (2.96% max.) values and the disrupted nature of the fuel itself. Criticality analysis of these fuel canisters has been performed over the years to reflect conditions expected during transit from the reactor to the Idaho National Engineering and Environmental Laboratory, water pool storage,1 and transport/dry-pack storage at Idaho Nuclear Technology and Engineering Center.2,3 None of these prior analyses reflect the potential disposal conditions for this fuel inside a postclosure repository.

  15. Temperature field due to time-dependent heat sources in a large rectangular grid - Derivation of analytical solution

    Energy Technology Data Exchange (ETDEWEB)

    Claesson, J.; Probert, T. [Lund Univ. (Sweden). Dept. of Building Physics and Mathematical Physics

    1996-01-01

    The temperature field in rock due to a large rectangular grid of heat releasing canisters containing nuclear waste is studied. The solution is by superposition divided into different parts. There is a global temperature field due to the large rectangular canister area, while a local field accounts for the remaining heat source problem. The global field is reduced to a single integral. The local field is also solved analytically using solutions for a finite line heat source and for an infinite grid of point sources. The local solution is reduced to three parts, each of which depends on two spatial coordinates only. The temperatures at the envelope of a canister are given by a single thermal resistance, which is given by an explicit formula. The results are illustrated by a few numerical examples dealing with the KBS-3 concept for storage of nuclear waste. 8 refs.

  16. Human factors analysis and design methods for nuclear waste retrieval systems: Human factors design methodology and integration plan

    Science.gov (United States)

    Casey, S. M.

    1980-06-01

    The nuclear waste retrieval system intended to be used for the removal of storage canisters (each canister containing a spent fuel rod assembly) located in an underground salt bed depository is discussed. The implementation of human factors engineering principles during the design and construction of the retrieval system facilities and equipment is reported. The methodology is structured around a basic system development effort involving preliminary development, equipment development, personnel subsystem development, and operational test and evaluation. Examples of application of the techniques in the analysis of human tasks, and equipment required in the removal of spent fuel canisters is provided. The framework for integrating human engineering with the rest of the system development effort is documented.

  17. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between April 2009 and September 2009

    Energy Technology Data Exchange (ETDEWEB)

    Haycox, Jon; Pettitt, Will [ASC, Applied Seismology Consultants, Shrewsbury, Shropshire (United Kingdom)

    2009-12-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and changes in pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation (Pettitt et al. 1999), and during stages of canister heating and tunnel pressurisation (Haycox and Pettitt 2005a, b, 2006a, b, Zolezzi et al. 2007, 2008, Duckworth et al. 2008, 2009, Haycox and Duckworth 2009). Further information on the previous monitoring undertaken can be found in Appendix 1. This report covers the period between 1st April 2009 and 30th September 2009 and is the ninth 6-monthly processing and interpretation of the results from the experiment.

  18. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between October 2008 and March 2009

    Energy Technology Data Exchange (ETDEWEB)

    Haycox, Jon; Duckworth, Damion [ASC, Applied Seismology Consultants, Shrewsbury, Shropshire (United Kingdom)

    2009-06-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and changes in pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation (Pettitt et al. 1999), and during stages of canister heating and tunnel pressurisation (Haycox and Pettitt 2005a, b, 2006a, b, Zolezzi et al. 2007, 2008, Duckworth et al. 2008, 2009). Further information on the previous monitoring undertaken can be found in Appendix 1. This report covers the period between 1st October 2008 and 31st March 2009 and is the eighth 6-monthly processing and interpretation of the results from the experiment.

  19. Aespoe Hard Rock Laboratory. Prototype Repository. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between April 2008 and September 2008

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, D.; Haycox, J.; Pettitt, W.S. (Applied Seismology Consultants, Shrewsbury (United Kingdom))

    2009-03-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation [Pettitt et al., 1999], and during stages of canister heating and tunnel pressurisation [Haycox et al., 2005a and 2005b; Haycox et al., 2006a and 2006b; Zolezzi et al., 2007 and Duckworth et al., 2008]. Further information on this monitoring can be found in Appendix I. This report covers the period between 1st April 2008 and 30th September 2008 and is the seventh instalment of the 6-monthly processing and interpretation of the results from the experiment.

  20. Influence of void ratio on phase change of thermal energy storage for heat pipe receiver

    Directory of Open Access Journals (Sweden)

    Xiaohong Gui

    2015-01-01

    Full Text Available In this paper, influence of void ratio on phase change of thermal storage unit for heat pipe receiver under microgravity is numerically simulated. Accordingly, mathematical model is set up. A solidification-melting model upon the enthalpy-porosity method is specially provided to deal with phase changes. The liquid fraction distribution of thermal storage unit of heat pipe receiver is shown. The fluctuation of melting ratio in PCM canister is indicated. Numerical results are compared with experimental ones in Japan. The results show that void cavity prevents the process of phase change greatly. PCM melts slowly during sunlight periods and freezes slowly during eclipse periods as void ratio increases. The utility ratio of PCM during both sunlight periods and eclipse periods decreases obviously with the improvement of void ratio. The thermal resistance of void cavity is much higher than that of PCM canister wall. Void cavity prevents the heat transfer between PCM zone and canister wall.