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Sample records for canister overpack mco

  1. Multi-Canister Overpack (MCO) Design Report

    International Nuclear Information System (INIS)

    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. Multi-Canister Overpack (MCO) Topical Report

    International Nuclear Information System (INIS)

    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

  3. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building

    International Nuclear Information System (INIS)

    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

  4. Potential Multi-Canister Overpack (MCO) Cask Drop in the K West Basin South Loadout Pit

    International Nuclear Information System (INIS)

    This calculation note documents the probabilistic calculation of a potential drop of a multi-canister overpack (MCO) cask or MCO cask and immersion pail at the K West Basin south loadout pit. The calculations are in support of the cask loading system (CLS) subproject alignment of CLS equipment in the K West Basin south loadout pit

  5. Warehouse Plan for the Multi Canister Overpacks (MCO) and Baskets

    International Nuclear Information System (INIS)

    The Multi-Canister Overpacks (MCOs) will contain spent nuclear fuel (SNF) removed from the K East and West Basins. The SNF will be placed in fuel storage baskets that will be stacked inside the MCOs. Approximately 400 MCOS and 2170 baskets will fabricated for this purpose. These MCOs, loaded with SNF, will be placed in interim storage in the Canister Storage Building (CSB) located in the 200 Area of the Hanford Site

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB) MULTI CANISTER OVERPACK (MCO) SAMPLING SYSTEM VALIDATION (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.; KLEM, M.J.

    2003-11-17

    Approximately 400 Multi-canister overpacks (MCO) containing spent nuclear fuel are to be interim stored at the Canister Storage Building (CSB). Several MCOs (monitored MCOs) are designated to be gas sampled periodically at the CSB sampling/weld station (Bader 2002a). The monitoring program includes pressure, temperature and gas composition measurements of monitored MCOs during their first two years of interim storage at the CSB. The MCO sample cart (CART-001) is used at the sampling/weld station to measure the monitored MCO gas temperature and pressure, obtain gas samples for laboratory analysis and refill the monitored MCO with high purity helium as needed. The sample cart and support equipment were functionally and operationally tested and validated before sampling of the first monitored MCO (H-036). This report documents the results of validation testing using training MCO (TR-003) at the CSB. Another report (Bader 2002b) documents the sample results from gas sampling of the first monitored MCO (H-036). Validation testing of the MCO gas sampling system showed the equipment and procedure as originally constituted will satisfactorily sample the first monitored MCO. Subsequent system and procedural improvements will provide increased flexibility and reliability for future MCO gas sampling. The physical operation of the sampling equipment during testing provided evidence that theoretical correlation factors for extrapolating MCO gas composition from sample results are unnecessarily conservative. Empirically derived correlation factors showed adequate conservatism and support use of the sample system for ongoing monitored MCO sampling.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 document SNF-6248

  14. Multi Canister Overpack (MCO) Handling Machine Trolley Seismic Uplift Constraint Design Loads

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-03-09

    The MCO Handling Machine (MHM) trolley moves along the top of the MHM bridge girders on east-west oriented rails. To prevent trolley wheel uplift during a seismic event, passive uplift constraints are provided as shown in Figure 1-1. North-south trolley wheel movement is prevented by flanges on the trolley wheels. When the MHM is positioned over a Multi-Canister Overpack (MCO) storage tube, east-west seismic restraints are activated to prevent trolley movement during MCO handling. The active seismic constraints consist of a plunger, which is inserted into slots positioned along the tracks as shown in Figure 1-1. When the MHM trolley is moving between storage tube positions, the active seismic restraints are not engaged. The MHM has been designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis (Reference 3) reported seismic uplift restraint loading and EDERER performed corresponding structural calculations. The ALSTHOM and EDERER calculations were performed with the east-west seismic restraints activated and the uplift restraints experiencing only vertical loading. 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 east-west trolley restraints are not engaged. For this case, the associated trolley movements would result in east-west lateral loads on the uplift constraints due to friction, as shown in Figure 1-2. During preliminary evaluations, questions were raised as to whether the EDERER calculations considered the latest ALSTHOM seismic analysis loads (See NCR No. 00-SNFP-0008, Reference 5). Further evaluation led to the conclusion that the EDERER calculations used appropriate vertical loading, but the uplift restraints would need to be re-analyzed and modified to account for lateral loading. The disposition of NCR 00-SNFP-0008 will track the redesign and modification effort. The purpose of this calculation is to establish bounding seismic

  15. Simulation of Multi Canister Overpack (MCO) Handling Machine Impact with Cask and MCO During Insertion into the Transfer Pit (FDT-137)

    International Nuclear Information System (INIS)

    The K-Basin Cask and Transportation System will be used for safely packaging and transporting approximately 2,100 metric tons of unprocessed, spent nuclear fuel from the 105 K East and K West Basins to the 200 E Area Canister Storage Building (CSB). Portions of the system will also be used for drying the spent fuel under cold vacuum conditions prior to placement in interim storage. The spent nuclear fuel is currently stored underwater in the two K-Basins. The K-Basins loadout pit is the area selected for loading spent nuclear fuel into the Multi-Canister Overpack (MCO) which in turn is located within the transportation cask. This Cask/MCO unit is secured.in the pit with a pail load out structure whose primary function is lo suspend and support the Cask/MCO unit at.the desired elevations and to protect the unit from the contaminated K-Basin water. The fuel elements will be placed in special baskets and stacked in the MCO that have been previously placed in the cask. The casks will be removed from the K Basin load out areas and taken to the cold vacuum drying station. Then the cask will be prepared for transportation to the CSB. The shipments will occur exclusively on the Hanford Site between K-Basins and the CSB. Travel will be by road with one cask per trailer. At the CSB receiving area the cask will be removed from the trailer. A gantry crane will then move the cask over to the transfer pit and load the cask into the transfer pit. From the transfer pit the MCO will be removed from the cask by the MCO Handling Machine (MHM). The MHM will move the MCO from the transfer pit to a canister storage tube in the CSB. MCOs will be piled two high in each canister Storage tube

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  17. Accident and Off-Normal Response and Recovery from Multi-Canister Overpack (MCO) Processing Events

    International Nuclear Information System (INIS)

    In the process of removing spent nuclear fuel (SNF) from the K Basins through its subsequent packaging, drymg, transportation and storage steps, the SNF Project must be able to respond to all anticipated or foreseeable off-normal and accident events that may occur. Response procedures and recovery plans need to be in place, personnel training established and implemented to ensure the project will be capable of appropriate actions. To establish suitable project planning, these events must first be identified and analyzed for their expected impact to the project. This document assesses all off-normal and accident events for their potential cross-facility or Multi-Canister Overpack (MCO) process reversal impact. Table 1 provides the methodology for establishing the event planning level and these events are provided in Table 2 along with the general response and recovery planning. Accidents and off-normal events of the SNF Project have been evaluated and are identified in the appropriate facility Safety Analysis Report (SAR) or in the transportation Safety Analysis Report for Packaging (SARP). Hazards and accidents are summarized from these safety analyses and listed in separate tables for each facility and the transportation system in Appendix A, along with identified off-normal events. The tables identify the general response time required to ensure a stable state after the event, governing response documents, and the events with potential cross-facility or SNF process reversal impacts. The event closure is predicated on stable state response time, impact to operations and the mitigated annual occurrence frequency of the event as developed in the hazard analysis process

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  1. DESIGN OF THE HANFORD MULTI CANISTER OVERPACK (MCO) and DEVELOPMENT and QUALIFICATION OF THE CLOSURE WELDING PROCESS

    International Nuclear Information System (INIS)

    Processing more than 2,100 metric tons of metallic uranium spent nuclear fuel (SNF) into large stainless steel containers called Multi-Canister Overpacks (MCOs) is one of the top priorities for the Department of Energy (DOE) at the Hanford Site, located in southeastern Washington state. The MCOs will be temporarily stored on site and eventually shipped to the federal geologic repository for long-term storage. MCOs are constructed and ''N''stamped in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Section III, Division 1, Class 1 Components. Final closure welding poses a challenge after the fuel is loaded. Performing required examination and testing activities (volumetric examination and hydrostatic leak testing) can be difficult, if not impractical. An ASME Code Case N-595-3, was written specifically to allow code stamping by addressing such closures and providing alternative rules. MCOs are the first SNF canisters within the DOE complex to successfully use this code case for receiving ASME stamps. This paper discusses the design of the MCO, application of the N-595-3 code case, and development and qualification of the final welded closure. The MCO design considers internal pressure and handling loads, as well as processing and interim storage activities. The MCO functions as the primary or innermost containment as part of an overall transportation package so the design also considered interface features with secondary and transport containers. The MCO, approximately 2 feet in diameter and nearly 14 feet tall, is constructed primarily of Type 304/304L stainless steel and the final pressure boundary is of all-welded construction. The closure-weld is made with the Gas Tungsten Arc Welding (GTAW) process, using an automatic, machine-welding mode. Examination and testing of the closure includes the N-595-3 specified requirements-progressive Liquid Penetrant testing (PT) and final helium leak testing. At completion of the closure

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Analysis for Spent Nuclear Fuel Multi-Canister Overpack (MCO) Drop into the Cask from the Multi-Canister Overpack - Handling Machine (MHM) with Air Cushion

    International Nuclear Information System (INIS)

    The purpose of this report is to investigate the potential for damage to the MCO during impact from an accidental drop from the MHM into the shipping cask. The MCO is dropped from a height of 8.2 feet above the cask enters the cask concentrically and falls the additional 12.83 feet to the cask bottom. Because of the interface fit between the MCO and the cask and the air entrapment the MCO fall velocity is slowed. The shipping cask is resting on an impact absorber at the time of impact. The energy absorbing properties of the impact absorber are included in this analysis

  4. Multi-Canister overpack dual pressure rating; TOPICAL

    International Nuclear Information System (INIS)

    The SNF Project will change the Multi-Canister Overpack (MCO) design pressure rating in the mechanical closure configuration to 150 psig to permit substitution of 304L/304 stainless steel for the higher cost XM-19 in the MCO collar. The 450 psig pressure rating for the final welded MCO will remain unchanged

  5. Multi-Canister overpack pressure testing

    International Nuclear Information System (INIS)

    The Multi-Canister Overpack (MCO) shield plug closure assembly will be hydrostatically tested at the fabricator's shop to the 150 psig design test requirement in accordance with the ASME Code. Additionally, the MCO shell and collar will be hydrostatically tested at the fabricator's shop to the 450 psig design test requirement. Commercial practice has not required a pressure test of the closure weld after spent fuel is loaded in the containers. Based on this precedent and Code Case N-595-I, the MCO closure weld will not be pressure tested in the field

  6. Multi-canister overpack: additional NRC requirements

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) established in the K Basin Spent Fuel Project, Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel Project (SNFP) facilities to achieve ''nuclear safety equivalency'' to comparable U.S. Nuclear Regulatory Commission licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Conditioning Facility or K Basins Path Forward Projects, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNFP facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements, to establish nuclear safety equivalency for the MCO

  7. Multi-Canister overpack internal HEPA filters

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, K.E.

    1998-11-03

    The rationale for locating a filter assembly inside each Multi-Canister Overpack (MCO) rather than include the filter in the Cold Vacuum Drying (CVD) process piping system was to eliminate the potential for contamination to the operators, processing equipment, and the MCO. The internal HEPA filters provide essential protection to facility workers from alpha contamination, both external skin contamination and potential internal depositions. Filters installed in the CVD process piping cannot mitigate potential contamination when breaking the process piping connections. Experience with K-Basin material has shown that even an extremely small release can result in personnel contamination and costly schedule disruptions to perform equipment and facility decontamination. Incorporating the filter function internal to the MCO rather than external is consistent with ALARA requirements of 10 CFR 835. Based on the above, the SNF Project position is to retain the internal HEPA filters in the MCO design.

  8. Drop Testing Representative Multi-Canister Overpacks

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Criticality Safety Evaluation Report for the Multi-Canister Overpack

    International Nuclear Information System (INIS)

    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 Keff = 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

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

    International Nuclear Information System (INIS)

    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 keff = 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

  11. Application of transient ignition model to multi-canister (MCO) accident analysis

    International Nuclear Information System (INIS)

    The potential for ignition of spent nuclear fuel in a Multi-Canister Overpack (MCO) is examined. A transient model is applied to calculate the highest ambient gas temperature outside an MCO wall tube or shipping cask for which a stable temperature condition exists. This integral analysis couples reaction kinetics with a description of the MCO configuration, heat and mass transfer, and fission product phenomena. It thereby allows ignition theory to be applied to various complex scenarios, including MCO water loss accidents and dry MCO air ingression

  12. Multi-canister overpack closure operations location study

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel Path Forward Project (SNF Project) has been established to develop engineered methods for the expedited removal of the irradiated uranium fuel from the K East (KE) and K West (KW) Basins. As specified by the SNF Project, the SNF will be removed from the K Basins, conditioned for dry storage and placed in a long term interim storage facility located in the 200 East Area. The SNF primarily consists of Zircaloy-2 clad uranium fuel discharged from the N-Reactor. A small portion of the SNF is Single Pass Reactor (SPR) Fuel, which is aluminum clad uranium fuel. The SNF will be loaded into Multi-Canister Overpacks (MCOs) at the K Basins, transferred to the Cold Vacuum Drying (CVD) facility for initial fuel conditioning, and transported to the Canister Storage Building (CSB) for staging, final fuel conditioning, and dry storage. The MCO is a transportation, conditioning, and storage vessel. The MCO consists of a 24 inch pipe with a welded bottom closure and a top closure that is field welded after the MCO is loaded with SNF. The MCO is handled and transported in the vertical orientation during all operations. Except for operations within the CSB, the MCO is always within the transportation cask which primarily provides radiological shielding and structural protection of the MCO. The MCO closure operations location study provides a relative evaluation of location options at the K Basins and the CVD Facility and recommends that the MCO closure weld be performed, inspected, and repaired at the CVD Facility

  13. Criticality safety evaluation report for the multi-canister overpack; TOPICAL

    International Nuclear Information System (INIS)

    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

  14. Multi-canister overpack operations and maintenance manual

    International Nuclear Information System (INIS)

    This manual provides general operating and maintenance instructions for the Multi-Canister Overpack. Procedure outlines included are conceptual in nature and will be modified, expanded, and refined during preparation of detailed operating procedures

  15. Multi-canister overpack design report

    International Nuclear Information System (INIS)

    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. Multi-canister overpack design report

    Energy Technology Data Exchange (ETDEWEB)

    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 .

  17. Warehouse Plan for the Multi-Canister Overpacks (MC0) and Baskets

    International Nuclear Information System (INIS)

    The Multi-Canister Overpacks (MCO) will contain spent nuclear fuel (SNF) removed from the K East and West Basins. The SNF will be placed in fuel storage baskets that will be stacked inside the MCOs. Approximately 400 MCOs and 21 70 baskets will be fabricated for this purpose. These MCOs, loaded with SNF, will be placed in interim storage in the Canister Storage Building (CSB) located in the 200 Area of the Hanford Site. The MCOs consist of different components/sub-assemblies that will be manufactured by one or more vendors. All component/sub-assemblies will be shipped to the Hanford Site Central Stores Warehouse, 2355 Stevens Drive, Building 1163 in the 1100 Area, for inspection and storage until these components are required at the CSB and K Basins. The MCO fuel storage baskets will be manufactured in the MCO basket fabrication shop located in Building 328 of the Hanford Site 300 Area. The MCO baskets will be inspected at the fabrication shop before shipment to the Central Stores Warehouse for storage. The MCO components and baskets will be stored as received from the manufacturer with specified protective coatings, wrappings, and packaging intact to maintain mechanical integrity of the components and to prevent corrosion. The components and baskets will be shipped as needed from the warehouse to the CSB and K Basins. This warehouse plan includes the requirements for receipt of MCO components and baskets from the manufacturers and storage at the Hanford Site Central Stores Warehouse. Transportation of the MCO components and baskets from the warehouse, unwrapping, and assembly of the MCOs are the responsibility of SNF Operations and are not included in this plan

  18. Spent nuclear fuel project multi-canister overpack, additional NRC requirements

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE), established in the K Basin Spent Nuclear Fuel Project Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel (SNF) Project facilities to achieve nuclear safety equivalency to comparable US Nuclear Regulatory Commission (NRC)-licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Cold Vacuum Drying (CVD) facility or Hot Conditioning System, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNF Project facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements needed, in combination with the existing and applicable DOE requirements, to establish nuclear safety equivalency for the MCO. The background, basic safety issues and general comparison of NRC and DOE requirements for the SNF Project are presented in WHC-SD-SNF-DB-002

  19. Multi-Canister overpack design pressure rating

    International Nuclear Information System (INIS)

    The SNF project was directed to increase the MCO pressure rating by the U.S. Department of Energy, Richland Operations Office (RL) unless the action was shown to be cost prohibitive. This guidance was driven by RL's assessment that there was a need to improve margin and reduce risks associated with assumptions supporting the bounding pressure calculation for the MCO Sealing Strategy. Although more recent pressure analyses show a bounding MCO pressure of 50 psig, RL still considers it prudent to retain the pressure margin the 450 psig rating provides. This rating creates a real, clearly definable margin and significantly reduces the risk that the safety basis will be challenged

  20. Hanford Spent Nuclear Fuel Project evaluation of multi-canister overpack venting and monitoring options during staging of K basins fuel

    International Nuclear Information System (INIS)

    This engineering study recommends whether multi-canister overpacks containing spent nuclear fuel from the Hanford K Basins should be staged in vented or a sealed, but ventable, condition during staging at the Canister Storage Building prior to hot vacuum conditioning and interim storage. The integrally related issues of MCO monitoring, end point criteria, and assessing the practicality of avoiding venting and Hot Vacuum Conditioning for a portion of the spent fuel are also considered

  1. Feasibility of direct reactivity measurement in multi-canister overpacks at the Cold Vacuum Drying Facility

    International Nuclear Information System (INIS)

    A proposed method for measuring the chemical reaction rate (power) of breached N-Reactor fuel elements with water in a Multi-canister overpack (MCO) based on hydrogen release rate is evaluated. The reaction rate is measured at 50 C in an oxygen free water by applying a vacuum to boil the water and adding a low, measured flow of helium. The ratio of helium to hydrogen is used to infer the reaction rate. A test duration of less than 8 hours was found to provide sufficient accuracy for confidence in the measurement results. A more rigorous treatment of system measurement accuracy, which may yield shorter test durations, should be performed if this reactivity measurement is to be employed

  2. Estimates of particulate mass in multi-canister overpacks

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Sensitivity of probabilistic MCO water content estimates to key assumptions

    International Nuclear Information System (INIS)

    Sensitivity of probabilistic multi-canister overpack (MCO) water content estimates to key assumptions is evaluated with emphasis on the largest non-cladding film-contributors, water borne by particulates adhering to damage sites, and water borne by canister particulate. Calculations considered different choices of damage state degree of independence, different choices of percentile for reference high inputs, three types of input probability density function (pdfs): triangular, log-normal, and Weibull, and the number of scrap baskets in an MCO

  4. MCO loading and cask loadout technical manual

    International Nuclear Information System (INIS)

    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

  5. MCO loading and cask loadout technical manual

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  7. As-Built Verification Plan Spent Nuclear Fuel Canister Storage Building MCO Handling Machine

    International Nuclear Information System (INIS)

    This as-built verification plan outlines the methodology and responsibilities that will be implemented during the as-built field verification activity for the Canister Storage Building (CSB) MCO HANDLING MACHINE (MHM). This as-built verification plan covers THE ELECTRICAL PORTION of the CONSTRUCTION PERFORMED BY POWER CITY UNDER CONTRACT TO MOWAT. The as-built verifications will be performed in accordance Administrative Procedure AP 6-012-00, Spent Nuclear Fuel Project As-Built Verification Plan Development Process, revision I. The results of the verification walkdown will be documented in a verification walkdown completion package, approved by the Design Authority (DA), and maintained in the CSB project files

  8. Sandia studies of high-level waste canisters and overpacks applicable for a salt repository

    International Nuclear Information System (INIS)

    An experimental program to develop candidate materials for use as high-level waste (HLW) overpacks or canisters in a salt repository has been in progress at Sandia National Laboratories since 1976. The main objective of this program is to provide a waste package barrier having a long lifetime in the chemical and physical environment of a repository. This paper summarizes the recent corrosion and metallurgical study results for the prime overpack material, TiCode-12, in the areas of uniform corrosion (extremely low rate and extent); local attack, e.g., pits and crevices (none were found); stress corrosion cracking susceptibility (no significant changes in macroscopic tensile properties were detected); hydrogen sorption-embrittlement effects; effects of gamma irradiation in solution; and sensitization effects (testing is still in process in the last three areas). Previous candidate screening analyses on other alloys and recent work on alternate overpack alloys are reviewed. All phases of these interrelated laboratory, hot-cell, and field experimental studies are described. 16 references, 8 figures, 4 tables

  9. Multi-canister overpack project - verification and validation, MCNP 4A

    International Nuclear Information System (INIS)

    This supporting document contains the software verification and validation (V and V) package used for Phase 2 design of the Spent Nuclear Fuel Multi-Canister Overpack. V and V packages for both ANSYS and MCNP are included. Description of Verification Run(s): This software requires that it be compiled specifically for the machine it is to be used on. Therefore to facilitate ease in the verification process the software automatically runs 25 sample problems to ensure proper installation and compilation. Once the runs are completed the software checks for verification by performing a file comparison on the new output file and the old output file. Any differences between any of the files will cause a verification error. Due to the manner in which the verification is completed a verification error does not necessarily indicate a problem. This indicates that a closer look at the output files is needed to determine the cause of the error

  10. Impact of Aluminum on Anticipated Corrosion in a Flooded spent nuclear fuel Multi -Canister Overpack

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  13. SNF project's MCO compliance assessment with DOE ''general design criteria,'' order 6430.1A and ''SNF project MCO additional NRC requirements,'' HNF-SD-SNF-DB-005

    International Nuclear Information System (INIS)

    This document is presented to demonstrate the MCOs compliance to the major design criteria invoked on the MCO. This document is broken down into a section for the MCO's evaluation against DOE Order 6430.1A General Design Criteria sixteen divisions and then the evaluation of the MCO against HNF-SD-SNF-DB-005 ''Spent Nuclear Fuel Project Multi-Canister Overpack Additional NRC Requirements.'' The compliance assessment is presented as a matrix in tabular form. The MCO is the primary container for the K-basin's spent nuclear fuel as it leaves the basin pools and through to the 40 year interim storage at the Canister Storage Building (CSB). The MCO and its components interface with; the K basins, shipping cask and transportation system, Cold Vacuum Drying facility individual process bays and equipment, and CSB facility including the MCO handling machine (MHM), the storage tubes, and the MCO work stations where sampling, welding, and inspection of the MCO is performed. As the MCO is the primary boundary for handling, process, and storage, its main goals are to minimize the spread of its radiological contents to the outside of the MCO and provide for nuclear criticality control. The MCO contains personnel radiation shielding only on its upper end, in the form of a shield plug, where the process interfaces are located. Shielding beyond the shield plug is the responsibility of the using facilities. The design of the MCO and its components is depicted in drawings H-2-828040 through H-2-828075. Not every drawing number in the sequence is used. The first drawing number, H-2-828040, is the drawing index for the MCO. The design performance specification for the MCO is HW-S-0426, and was reviewed and approved by the interfacing design authorities, the safety, regulatory, and operations groups, and the local DOE office. The current revision for the design performance specification is revision 5. The designs of the MCO have been reviewed and approved in a similar way and the reports

  14. Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report - Project A.5 and A.6

    International Nuclear Information System (INIS)

    This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01

  15. Probability analysis of MCO over-pressurization during staging

    International Nuclear Information System (INIS)

    The purpose of this calculation is to determine the probability of Multi-Canister Overpacks (MCOs) over-pressurizing during staging at the Canister Storage Building (CSB). Pressurization of an MCO during staging is dependent upon changes to the MCO gas temperature and the build-up of reaction products during the staging period. These effects are predominantly limited by the amount of water that remains in the MCO following cold vacuum drying that is available for reaction during staging conditions. Because of the potential for increased pressure within an MCO, provisions for a filtered pressure relief valve and rupture disk have been incorporated into the MCO design. This calculation provides an estimate of the frequency that an MCO will contain enough water to pressurize beyond the limits of these design features. The results of this calculation will be used in support of further safety analyses and operational planning efforts. Under the bounding steady state CSB condition assumed for this analysis, an MCO must contain less than 1.6 kg (3.7 lbm) of water available for reaction to preclude actuation of the pressure relief valve at 100 psid. To preclude actuation of the MCO rupture disk at 150 psid, an MCO must contain less than 2.5 kg (5.5 lbm) of water available for reaction. These limits are based on the assumption that hydrogen generated by uranium-water reactions is the sole source of gas produced within the MCO and that hydrates in fuel particulate are the primary source of water available for reactions during staging conditions. The results of this analysis conclude that the probability of the hydrate water content of an MCO exceeding 1.6 kg is 0.08 and the probability that it will exceed 2.5 kg is 0.01. This implies that approximately 32 of 400 staged MCOs may experience pressurization to the point where the pressure relief valve actuates. In the event that an MCO pressure relief valve fails to open, the probability is 1 in 100 that the MCO would experience

  16. Overhead Vertical Strike Analysis for the MCO in the CSB

    International Nuclear Information System (INIS)

    The purpose of this calculation is to document the structural adequacy of Multi-Canister Overpacks (MCOs) in two separate normal configurations for the 40-year interim period in the storage tubes at the Canister Storage Building (CSB). The two configurations apply when the lower MCO is arranged with (1) just a shield plug mechanical closure or (2) with the canister cover welded over the shield plug. Multiple analyses have been conducted to show the MCO design is adequate for drop events involving a lower, passive MCO being struck by an overhead drop load MCO both in vertical position. Minimal plastic deformations are experienced for either the MCO with shield plug or the MCO with cover cap. This survivability at 35 g's demonstrates the package's worthiness for 40 years of storage. The force between the two MCOs created by a drop event far exceed that resulting from stacking two loaded MCOs. The stacking load condition is not a structural concern and is, therefore, acceptable for the design duration of 40 years of storage. Both mechanical and welded cap models have the lifting ring and associated flat plate for the top unit to rest on and, therefore, easily fit together. Both designs were analyzed for the top unit to be dropped onto the lower unit. The mechanical seal unit only has high localized stress. The capped unit has a tendency to buckle but is still acceptable. Both designs will still be adequate for leakage and pressure during passive storage and are acceptable for the design duration of 40 years

  17. MCO combustible gas management leak test acceptance criteria

    International Nuclear Information System (INIS)

    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 10-5 std cc/sec and 1 x 10-7 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)

  18. Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report Project A.5 and A.6

    Energy Technology Data Exchange (ETDEWEB)

    ARD, K.E.

    2000-04-19

    This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  2. Packaging design criteria for the MCO cask

    International Nuclear Information System (INIS)

    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

  3. Packaging Design Criteria for the MCO Cask

    Energy Technology Data Exchange (ETDEWEB)

    FLANAGAN, B.D.

    2000-08-01

    Approximately 2,100 metric tons of unprocessed, irradiated, nuclear fuel elements are presently stored in the K Basins (including approximately 700 additional elements from the Plutonium-Uranium Extraction Plant, N Reactor, and 327 Laboratory). To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) in the 200 East Area. 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 Multi-canister Overpacks. Concurrent with the K Basin cleanup, 72 Shippingport Pressurized Water Reactor Core 2 fuel assemblies will be transported from T Plant to the CSB to provide space at T Plant for K Basin sludge canisters.

  4. Multicanister overpack topical report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Packaging design criteria for the MCO cask

    International Nuclear Information System (INIS)

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins (including possibly 700 additional elements from PUREX, N Reactor, and 327 Laboratory). The basin water, particularly in the K East Basin, contains significant quantities of dissolved nuclear isotopes and radioactive fuel corrosion particles. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) in the 200 East area. In order to initiate K Basin cleanup on schedule, the two-year fuel-shipping campaign must begin by December 1997. The purpose of this packaging design criteria is to provide criteria for the design, fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multiple Canister Overpacks

  6. Dynamic Impact Analyses and Tests of Concrete Overpacks - 13638

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    Concrete cask is an option for spent nuclear fuel interim storage which is prevailingly used in US. A concrete cask usually consists of metallic canister which confines the spent nuclear fuel and concrete overpack. When the overpack undergoes a severe missile impact which might be caused by a tornado or an aircraft crash, it should sustain acceptable level of structural integrity so that its radiation shielding capability and the retrievability of canister are maintained. Missile impact against a concrete overpack involves two damage modes, local damage and global damage. Local damage of concrete is usually evaluated by empirical formulas while the global damage is evaluated by finite element analysis. In many cases, those two damage modes are evaluated separately. In this research, a series of numerical simulations are performed using finite element analysis to evaluate the global damage of concrete overpack as well as its local damage under high speed missile impact. We consider two types of concrete overpack, one with steel in-cased concrete without reinforcement and the other with partially-confined reinforced concrete. The numerical simulation results are compared with test results and it is shown that appropriate modeling of material failure is crucial in this analysis and the results are highly dependent on the choice of failure parameters. (authors)

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

    International Nuclear Information System (INIS)

    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 requirements are noted

  8. Spent Nuclear Fuel [SNF] Project Canister Storage Building [CSB] Final Safety Analysis Report [FSAR] Volume 1 [Section 1-3

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) established the Spent Nuclear Fuel (SNF) Project to address safety and environmental concerns associated with deteriorating SNF presently stored under water in the Hanford Site K Basins, which are located in the 100 K Area near the Columbia River. Recommendations for a series of projects to construct and operate systems and facilities to manage the safe removal and storage of K Basins fuel were made in WHC-EP-0830, Hanford Spent Nuclear Fuel Recommended Path Forward, and its subsequent update, WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel. The integrated process strategy recommendations include the following steps: (1) Fuel preparation activities at the K Basins, including removing the fuel elements from their K Basins canisters; separating fuel particulate from fuel elements and fuel fragments greater than 0.25 in. in any dimension; removing excess sludge from the fuel fragments by means of flushing, as necessary; and packaging the fuel into multi-canister overpacks (MCOs); (2) Transportation of MCOs loaded with SNF from K Basins to the Cold Vacuum Drying Facility (CVDF); (3) Removal of free water by draining and vacuum drying at the CVDF in the 100 K Area; (4) Dry shipment of fuel from the CVDF to the Canister Storage Building (CSB), a new facility in the 200 East Area; and (5) Interim storage of the MCOs in the CSB until a suitable long-term repository is established. In addition, the CSB can also store Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies in a modified MCO container called the Shippingport spent fuel canister. The Interim Storage Area has been established adjacent to the CSB for storage of other non-defense SNF in above-ground dry cask storage containers

  9. 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.

  10. Creep life simulations of EB welded copper overpack

    International Nuclear Information System (INIS)

    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

  11. Impact Analyses and Tests of Concrete Overpacks of Spent Nuclear Fuel Storage Casks

    International Nuclear Information System (INIS)

    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, 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

  12. MCO closure welding process parameter development and qualification

    International Nuclear Information System (INIS)

    One of the key elements in the SNF process is final closure of the MCO by welding. Fuel is loaded into the MCO (approximately 2 ft. in diameter and 13 ft. long) and a heavy shield plug is inserted into the top, creating a mechanical seal. The plug contains several process ports for various operations, including vacuum drying and inert-gas backfilling of the packaged fuel. When fully processed, the Canister Cover Assembly (CCA) is placed over the shield plug and final closure made by welding. The following reports the effort between the Amer Industrial Technology (AIT) and Fluor Hanford (FH) to develop and qualify the welding process for making the final closure--with primary emphasis on developing a set of robust parameters for deposition of the root pass. Work was carried out in three phases: (1) Initial welding process and equipment selection with subsequent field demonstration testing; (2) Development and qualification of a specific process technique and parameters; and (3) Validation of the process and parameters at the CSB under mock production conditions. This work establishes the process technique and parameters that provide a high level of confidence that acceptable MCO closure welds will be made on a consistent and repeatable basis

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

    International Nuclear Information System (INIS)

    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

  14. 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.

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

    International Nuclear Information System (INIS)

    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.)

  16. Recommendations for codes and standards to be used for design and fabrication of high level waste canister

    International Nuclear Information System (INIS)

    This study identifies codes, standards, and regulatory requirements for developing design criteria for high-level waste (HLW) canisters for commercial operation. It has been determined that the canister should be designed as a pressure vessel without provision for any overpressure protection type devices. It is recommended that the HLW canister be designed and fabricated to the requirements of the ASME Section III Code, Division 1 rules, for Code Class 3 components. Identification of other applicable industry and regulatory guides and standards are provided in this report. Requirements for the Design Specification are found in the ASME Section III Code. It is recommended that design verification be conducted principally with prototype testing which will encompass normal and accident service conditions during all phases of the canister life. Adequacy of existing quality assurance and licensing standards for the canister was investigated. One of the recommendations derived from this study is a requirement that the canister be N stamped. In addition, acceptance standards for the HLW waste should be established and the waste qualified to those standards before the canister is sealed. A preliminary investigation of use of an overpack for the canister has been made, and it is concluded that the use of an overpack, as an integral part of overall canister design, is undesirable, both from a design and economics standpoint. However, use of shipping cask liners and overpack type containers at the Federal repository may make the canister and HLW management safer and more cost effective. There are several possible concepts for canister closure design. These concepts can be adapted to the canister with or without an overpack. A remote seal weld closure is considered to be one of the most suitable closure methods; however, mechanical seals should also be investigated

  17. Performance of the SKB copper/steel canister

    International Nuclear Information System (INIS)

    The performance of the SKB copper/steel canister has been analyzed. The present knowledge of long-term function of the canister is summarized. Radionuclide release calculations for a reference failure scenario and the effect of some variations on release rates are shown. The Features, Events and Processes (FEPs) that are affecting the studied scenarios have been classified according to the 'Rock Engineering Systems' methodology as defined by SKB for the copper/steel canister. Radionuclide release rate is calculated for a reference failure scenario where a small hole in the weld of the outer copper overpack is assumed to exist at the time of deposition. The hole in the copper overpack is assumed to be of a constant size until the inner steel canister looses its mechanical integrity. The steel is assumed to maintain mechanical stability during 5000 years and after this time period the hole through the copper is assumed to be 0.1 m2, which translate to insignificant transport resistance from the canister wall. The release rates for C-14, Sr-90, I-129, Cs-137, Pu-239 and Am-241 are calculated for the reference failure scenario and for a number of variations. The variations include glaciation, only few of the Zircaloy tubes damaged, different canister filling materials, variations in sorption properties of the bentonite clay and different life-time of the inner steel canister. The performance of the canister and near-field, concerning the release rates of the studied radionuclides, is as expected, comparable to the release rates obtained in SKB 91. 11 refs, figs, tabs

  18. Feasibility study for a DOE research and production fuel multipurpose canister

    International Nuclear Information System (INIS)

    This is a report of the feasibility of multipurpose canisters for transporting, storing, and sing of Department of Energy research and production spent nuclear fuel. Six representative Department of Energy fuel assemblies were selected, and preconceptual canister designs were developed to accommodate these assemblies. The study considered physical interface, structural adequacy, criticality safety, shielding capability, thermal performance of the canisters, and fuel storage site infrastructure. The external envelope of the canisters was designed to fit within the overpack casks for commercial canisters being developed for the Department of Energy Office of Civilian Radioactive Waste Management. The budgetary cost of canisters to handle all fuel considered is estimated at $170.8M. One large conceptual boiling water reactor canister design, developed for the Office of Civilian Radioactive Waste Management, and two new canister designs can accommodate at least 85% of the volume of the Department of Energy fuel considered. Canister use minimizes public radiation exposure and is cost effective compared with bare fuel handling. Results suggest the need for additional study of issues affecting canister use and for conceptual design development of the three canisters

  19. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    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 the current design of the CHF and may not reflect the ongoing design evolution of the facility

  20. Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements

    International Nuclear Information System (INIS)

    The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel

  1. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    This document constitutes the final report from 'SKBs reference group for mechanical integrity of canisters for spent nuclear fuel'. A complete list of all reports initiated by the reference group can be found in the summary report in this document. The main task of the reference group has been to advice SKB regarding the choice (ranking of alternatives) of canister type for different types of storage. The choice should be based on requirements of impermeability for a given time period and identification of possible limiting mechanisms. The main conclusions from the work were: From mechanical point of view, low phosphorous oxygen free copper (Cu-OFP) is a preferred canisters material. It exhibits satisfactory ductility both during tensile and creep testing. The residual stresses in the canisters are of such a magnitude that the estimated time to creep rupture with the data obtained for the Cu-OFP material is essentially infinite. Based on the present knowledge of stress corrosion cracking of copper there appears to be a small risk for such to occur in the projected environment. This risk need some further study. Rock shear movements of the size of 10 cm should pose no direct threat to the integrity of the canisters. Considering mechanical integrity, the composite copper/steel canister is an advantageous alternative. The recommendations for further research included continued studies of the creep properties of copper and of stress corrosion cracking. However, the studies should focus more directly on the design and fabrication aspect of the canister

  2. The concrete canister program

    International Nuclear Information System (INIS)

    In the spring of 1974, WNRE began development and demonstration of a dry storage concept, called the concrete canister, as a possible alternative to storage of irradiated CANDU fuel in water pools. The canister is a thick-walled concrete monolith containing baskets of fuel in the dry state. The decay heat from the fuel is dissipated to the environment by natural heat transfer. Four canisters were designed and constructed. Two canisters containing electric heaters have been subjected to heat loads of 2.5 times the design, ramp heat-load cycling, and simulated weathering tests. The other two canisters were loaded with irradiated fuel, one containing fuel bundles of uniform decay heat and the other containing bundles of non-uniform decay heat in a non-symmetrical radial and axial array. The collected data were used to verify the analytical tools for prediction of effectiveness of heat transfer and radiation shielding and to verify the design of the basket and canisters. The demonstration canisters have shown that this concept is a viable alternative to water pools for the storage of irradiated CANDU fuel. (author)

  3. Drying behavior of K-East canister sludge

    International Nuclear Information System (INIS)

    A series of tests were conducted by Pacific Northwest National Laboratory to evaluate the drying behavior of sludge taken from the Hanford K-East Basin storage canisters. Some of the components of K-Basin sludge, such as oxides of uranium and its hydrates, could be associated with the spent nuclear fuel that will ultimately be loaded into Multi-Canister Overpacks (MCOs) and transferred to interim dry storage on the Hanford Site. The materials sealed in the MCOs must be compatible with the storage facility safety basis and the design accident analyses. Understanding the drying behavior of hydrates that may be formed by the reaction of uranium oxides (corrosion products) and water will help ensure these criteria are addressed. Drying measurements of sludge samples collected from K-East Basin canisters showed the water content (physically plus chemically bound) to range between 5 wt% and 75 wt%. Uranium oxide hydrates, the main source of gaseous products that can pressurize the MCOs during storage, constituted about 3 wt% to 15 wt% of the total water content of the initial weight. Most of the physically bound water was assumed to be released from the samples at ambient temperature when the system was pumped down to vacuum conditions of about 40 mTorr. The period for release of most free water in the K-East canister sludge was about 24 hours

  4. Design basis for the copper/steel canister. Stage four. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the points of view of choice of materials, manufacturing technology and quality assurance. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. Nodular (or ductile) iron is selected and this is capable of providing mechanical properties which are equally suitable as those of the originally selected high strength low alloy steel. The material specified for the overpack is Oxygen free copper with 50 ppm of phosphorus added. Corrosion studies supported by SKB indicate that in the absence of mechanical failure or accelerated localised corrosion the overpack should provide corrosion shielding of the canister for its full design life. Published work claiming that the nodular iron liner would have corrosion characteristics similar to the carbon steel which had been examined in depth is flawed since the microstructures of the iron and carbon steel specimens used were not investigated. It is highly unlikely that nodular irons in the form used for the experiments would have similar structures to nodular iron in the canisters by chance. If the overpack were breached during the aerobic period of the repository life then very rapid penetration of the inner liner could occur. It has been recognised that the roll forming method is not suitable for serial production and alternatives are being sought. The electron beam welding process has been explored with tenacity but has so far failed to produce a satisfactory lid weld. A new welder is being developed for supply to the SKB pilot plant where development will be continued. An alternative welding process, friction stir welding, is being examined as a candidate for attaching lids. Surface breaking defects may be detected using eddy current methods but there is currently no reliable way of detecting small sub surface defects in the overpack

  5. Design basis for the copper/steel canister. Stage four. Final report

    International Nuclear Information System (INIS)

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the points of view of choice of materials, manufacturing technology and quality assurance. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. Nodular (or ductile) iron is selected and this is capable of providing mechanical properties which are equally suitable as those of the originally selected high strength low alloy steel. The material specified for the overpack is Oxygen free copper with 50 ppm of phosphorus added. Corrosion studies supported by SKB indicate that in the absence of mechanical failure or accelerated localised corrosion the overpack should provide corrosion shielding of the canister for its full design life. Published work claiming that the nodular iron liner would have corrosion characteristics similar to the carbon steel which had been examined in depth is flawed since the microstructures of the iron and carbon steel specimens used were not investigated. It is highly unlikely that nodular irons in the form used for the experiments would have similar structures to nodular iron in the canisters by chance. If the overpack were breached during the aerobic period of the repository life then very rapid penetration of the inner liner could occur. It has been recognised that the roll forming method is not suitable for serial production and alternatives are being sought. The electron beam welding process has been explored with tenacity but has so far failed to produce a satisfactory lid weld. A new welder is being developed for supply to the SKB pilot plant where development will be continued. An alternative welding process, friction stir welding, is being examined as a candidate for attaching lids. Surface breaking defects may be detected using eddy current methods but there is currently no reliable way of detecting small sub surface defects in the overpack

  6. Interim Storage of RH-TRU 72B Canisters at the DOE Oak Ridge Reservation

    International Nuclear Information System (INIS)

    This paper describes an evaluation performed by the Department of Energy (DOE) Oak Ridge Operations (ORO) office for potential interim storage of remote-handled (RH) transuranic (TRU) 72B waste canisters at the Oak Ridge National Laboratory (ORNL). The evaluation included the conceptual design of a devoted canister storage facility and an assessment of the existing RHTRU waste storage facilities for storage of canisters. The concept for the devoted facility used modular concrete silos located on an above-grade storage pad. The assessment of the existing facilities considered the potential methods, facility modifications, and conceptual equipment that might be used for storage of 400 millisievert per hour (mSv/hr) canisters. The results of the evaluation indicated that the initial investment into a devoted facility was relatively high as compared to the certainty that significant storage capacity was necessary prior to the Waste Isolation Pilot Plant (WIPP) accepting RH-TRU waste for disposal. As an alternative, the use of individual concrete overpacks provided an incremental method that could be used with the existing storage facilities and outside storage pads. For the concrete overpack concepts considered, the cylindrical design stored in a vertical orientation was determined to be the most effective

  7. Mechanical failure of SKB spent fuel disposal canisters. Mathematical modelling and scoping calculations

    International Nuclear Information System (INIS)

    According to the current design of SKB, a copper overpack with a cast steel inner component will be used as the disposal canister for spent nuclear fuel. A recent study considered the case of a breach in the copper overpack, through which groundwater could enter the canister. It has pointed out that hydrogen gas generated by an anaerobic corrosion could cushion the system and reduce or eventually stop further infiltration of water into the breached canister, and thence the spent fuel. One potential pitfall in this previous study lies in the fact that it did not consider any processes which might violate the following assumptions which are essential for the gas 'cushioning': 1. Hydrogen gas accumulated in the annular gap in the canister forms a free gas phase which is stable indefinitely into future; 2. Elevated gas pressure in the canister prevents further supply of groundwater except for diffusion of vapour. In the current study we developed a set of mathematical models for the above problem and applied it to carry out an independent assessment of the long-term behaviour of the canister. A key aim in this study was to clarify whether there are any alternative processes which may affect the result obtained by the previous study by violating one of the assumptions listed above. For this purpose, a scenario development exercise was conducted. The result supported the concept described in the previous study. One exception is that possible intrusion of bentonite gel followed by its desaturation could leave paths both for the gas and water simultaneously without forming a gas cushion. This is summarised in the first part of the report. In the second part, development of mathematical models and their applications are described. The key results are: 1. The model describing behaviour of gas and pore water in the canister and the buffer material reproduced the main results of the previous study; 2. The model considering intrusion of the bentonite gel pointed out possibility

  8. Preparing, Loading and Shipping Irradiated Metals in Canisters Classified as Remote-Handled (RH) Low-Level Waste (LLW) From Oak Ridge National Laboratory (ORNL) to the Nevada Test Site (NTS)

    International Nuclear Information System (INIS)

    Irradiated metals, classified as remote-handled low-level waste generated at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, were containerised in various sized canisters for long-term storage. The legacy waste canisters were placed in below-grade wells located at the 7827 Facility until a pathway for final disposal at the Nevada Test Site (NTS) could be identified and approved. Once the pathway was approved, WESKEM, LLC was selected by Bechtel Jacobs Company, LLC to prepare, load, and ship these canisters from ORNL to the NTS. This paper details some of the technical challenges encountered during the retrieval process and solutions implemented to ensure the waste was safely and efficiently over-packed and shipped for final disposal. The technical challenges detailed in this paper include: 1) how to best perform canister/lanyard pre-lift inspections since some canisters had not been moved in ∼10 years, so deterioration was a concern; 2) replacing or removing damaged canister lanyards; 3) correcting a mis-cut waste canister lanyard resulting in a shielded overpack lid not seating properly; 4) retrieving a stuck canister; and 5) developing a path forward after an overstrained lanyard failed causing a well shield plug to fall and come in contact with a waste canister. Several of these methods can serve as positive lessons learned for other projects encountering similar situations. (authors)

  9. 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.

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

    International Nuclear Information System (INIS)

    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

  11. 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

  12. Design review report for the MCO loading system

    International Nuclear Information System (INIS)

    This design report presents the design of the MCO Loading System. The report includes final design drawings, a system description, failure modes and recovery plans, a system operational description, and stress analysis

  13. Improved Air-Treatment Canister

    Science.gov (United States)

    Boehm, A. M.

    1982-01-01

    Proposed air-treatment canister integrates a heater-in-tube water evaporator into canister header. Improved design prevents water from condensing and contaminating chemicals that regenerate the air. Heater is evenly spiraled about the inlet header on the canister. Evaporator is brazed to the header.

  14. Status report, canister fabrication

    International Nuclear Information System (INIS)

    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

  15. 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

  16. K West Basin canister survey

    International Nuclear Information System (INIS)

    A survey was conducted of the K West Basin to determine the distribution of canister types that contain the irradiated N Reactor fuel. An underwater camera was used to conduct the survey during June 1998, and the results were recorded on videotape. A full row-by-row survey of the entire basin was performed, with the distinction between aluminum and stainless steel Mark 1 canisters made by the presence or absence of steel rings on the canister trunions (aluminum canisters have the steel rings). The results of the survey are presented in tables and figures. Grid maps of the three bays show the canister lid ID number and the canister type in each location that contained fuel. The following abbreviations are used in the grid maps for canister type designation: IA = Mark 1 aluminum, IS = Mark 1 stainless steel, and 2 = Mark 2 stainless steel. An overall summary of the canister distribution survey is presented in Table 1. The total number of canisters found to contain fuel was 3842, with 20% being Mark 1 Al, 25% being Mark 1 SS, and 55% being Mark 2 SS. The aluminum canisters were predominantly located in the East and West bays of the basin

  17. Thermal Analysis of Cold Vacuum Drying (CVD) of Spent Nuclear Fuel (SNF)

    International Nuclear Information System (INIS)

    The thermal analysis examined transient thermal and chemical behavior of the Multi-Canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with N Reactor spent fuel. This analysis provides the basis for the MCO thermal behavior at the CVD Facility in support of the safety basis documentation

  18. Thermal Analysis of Cold Vacuum Drying (CVD) of Spent Nuclear Fuel (SNF)

    Energy Technology Data Exchange (ETDEWEB)

    PIEPHO, M.G.

    2000-03-23

    The thermal analysis examined transient thermal and chemical behavior of the Multi-Canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with N Reactor spent fuel. This analysis provides the basis for the MCO thermal behavior at the CVD Facility in support of the safety basis documentation.

  19. Drop Accidents in the Canister Storage Building (CSB) Addressed by Design Features and or Design Calculations

    International Nuclear Information System (INIS)

    A variety of drop shear or impact scenarios have been identified for the Canister Storage Building. Some of these are being addressed by new calculations or require no specific action. This document describes five of them which are addressed by design features and/or existing design calculations. For each of the five a position is stated indicating the reason for assurance that the safety functions of the MCO will not be jeopardized by the accident. Following the position is a description of the basis for that position

  20. MCO gas composition for low reactive surface areas

    International Nuclear Information System (INIS)

    This calculation adjusts modeled output (HNF-SD-SNF-TI-040, Rev. 2) by considering lower reactive fuel surface areas and by increasing the input helium backfill overpressure from 0.5 to 1.5 atm (2.5 atm abs) to verify that MCO gas-phase oxygen concentrations can remain below 4 mole % over a 40 year interim period under a worst case condition of zero reactive surface area. Added backfill gas will dilute any gases generated during interim storage and is a strategy within the current design capability. The zero reactive surface area represents a hypothetical worst case example where there is no fuel scrap and/or damaged spent fuel rods in an MCO. Also included is a hypothetical case where only K East fuel exists in an MCO with an added backfill overpressure of 0.5 atm (1.5 atm abs)

  1. MCO gas composition for low reactive surface areas

    Energy Technology Data Exchange (ETDEWEB)

    Packer, M.J.

    1998-07-23

    This calculation adjusts modeled output (HNF-SD-SNF-TI-040, Rev. 2) by considering lower reactive fuel surface areas and by increasing the input helium backfill overpressure from 0.5 to 1.5 atm (2.5 atm abs) to verify that MCO gas-phase oxygen concentrations can remain below 4 mole % over a 40 year interim period under a worst case condition of zero reactive surface area. Added backfill gas will dilute any gases generated during interim storage and is a strategy within the current design capability. The zero reactive surface area represents a hypothetical worst case example where there is no fuel scrap and/or damaged spent fuel rods in an MCO. Also included is a hypothetical case where only K East fuel exists in an MCO with an added backfill overpressure of 0.5 atm (1.5 atm abs).

  2. Trial manufacturing of titanium-carbon steel composite overpack

    Energy Technology Data Exchange (ETDEWEB)

    Honma, Nobuyuki; Chiba, Takahiko; Tanai, Kenji [Japan Nuclear Cycle Development Inst., Waste Management and Fuel Cycle Research Center, Tokai, Ibaraki (Japan)

    1999-11-01

    This paper reports the results of design analysis and trial manufacturing of full-scale titanium-carbon steel composite overpacks. The overpack is one of the key components of the engineered barrier system, hence, it is necessary to confirm the applicability of current technique in their manufacture. The required thickness was calculated according to mechanical resistance analysis, based on models used in current nuclear facilities. The Adequacy of the calculated dimensions was confirmed by finite-element methods. To investigate the necessity of a radiation shielding function of the overpack, the irradiation from vitrified waste has been calculated. As a result, it was shown that shielding on handling and transport equipment is a more reasonable and practical approach than to increase thickness of overpack to attain a self-shielding capability. After the above investigation, trial manufacturing of full-scale model of titanium-carbon steel composite overpack has been carried out. For corrosion-resistant material, ASTM Grade-2 titanium was selected. The titanium layer was bonded individually to a cylindrical shell and fiat cover plates (top and bottom) made of carbon steel. For the cylindrical shell portion, a cylindrically formed titanium layer was fitted to the inner carbon steel vessel by shrinkage. For the flat cover plates (top and bottom), titanium plate material was coated by explosive bonding. Electron beam welding and gas metal arc welding were combined to weld of the cover plates to the body. No significant failure was evident from inspections of the fabrication process, and the applicability of current technology for manufacturing titanium-carbon steel composite overpack was confirmed. Future research and development items regarding titanium-carbon steel composite overpacks are also discussed. (author)

  3. Trial manufacturing of titanium-carbon steel composite overpack

    International Nuclear Information System (INIS)

    This paper reports the results of design analysis and trial manufacturing of full-scale titanium-carbon steel composite overpacks. The overpack is one of the key components of the engineered barrier system, hence, it is necessary to confirm the applicability of current technique in their manufacture. The required thickness was calculated according to mechanical resistance analysis, based on models used in current nuclear facilities. The Adequacy of the calculated dimensions was confirmed by finite-element methods. To investigate the necessity of a radiation shielding function of the overpack, the irradiation from vitrified waste has been calculated. As a result, it was shown that shielding on handling and transport equipment is a more reasonable and practical approach than to increase thickness of overpack to attain a self-shielding capability. After the above investigation, trial manufacturing of full-scale model of titanium-carbon steel composite overpack has been carried out. For corrosion-resistant material, ASTM Grade-2 titanium was selected. The titanium layer was bonded individually to a cylindrical shell and fiat cover plates (top and bottom) made of carbon steel. For the cylindrical shell portion, a cylindrically formed titanium layer was fitted to the inner carbon steel vessel by shrinkage. For the flat cover plates (top and bottom), titanium plate material was coated by explosive bonding. Electron beam welding and gas metal arc welding were combined to weld of the cover plates to the body. No significant failure was evident from inspections of the fabrication process, and the applicability of current technology for manufacturing titanium-carbon steel composite overpack was confirmed. Future research and development items regarding titanium-carbon steel composite overpacks are also discussed. (author)

  4. Trial manufacturing of copper-carbon steel composite overpack

    International Nuclear Information System (INIS)

    This paper reports the results of design analysis and trial manufacturing of copper-carbon steel composite overpacks. The overpack is one of the key components of the engineered barrier system, hence, it is necessary to confirm the applicability of current technique in their manufacture. The copper-carbon steel composite overpack consists of a double container, an outer vessel made of oxygen-free, high-purity copper as the corrosion allowance material, and an inner vessel made of carbon steel as the pressure-resistant material. The trial manufacturing in this time, only the copper outer vessel has been fabricated. Both oxygen-free copper and oxygen-free phosphorus copper were used as materials for the outer vessel. For the shell and bottom portion, these materials were formed integrally by a backward extrusion method. For sealing the top cover plate to the main body, an electron-beam welding method was applied. After manufacturing, mechanical testing of specimens from the copper vessels were carried out. It was confirmed that current technique has sufficient feasibility to manufacture outer vessel. In addition, potential for irradiation embrittlement of the inner carbon-steel vessel by irradiation from vitrified waste over the life time of the overpack has been analyzed. It was shown that the small degree of irradiation embrittlement gives no significant impact on the pressure resistance of the carbon-steel vessel. Future research and development items regarding copper-carbon steel composite overpacks are also discussed. (author)

  5. Moisture insensitive charcoal canisters

    International Nuclear Information System (INIS)

    Continuous monitoring of 222Rn concentrations in the air in houses is the most appropriate approach for the real-time measurements, but this requires complex and expensive instruments and is not practical for large studies. Activated carbon canisters have been used extensively for determining the average concentration over a period of a few days. The ''open face'' charcoal detectors have an integration time constant of about 14 h so that they are sensitive to short-term transient changes in the radon concentration. In addition, water uptake at high relative humidities reduces the radon uptake by the charcoal. The addition of a diffusion barrier and a nylon screen results in a charcoal detector with an integration half-time ranging from 20 to 60 h and a reduced uptake of water at high humidities. Silicone rubber sheeting is relatively permeable to radon and impermeable to water vapor. It was the purpose of this study to evaluate the effect of a silicone barrier on the charcoal canister radon collective device. 3 refs

  6. Investigation of the thermal behavior of 2 1/2 ton cylinder protective overpack

    International Nuclear Information System (INIS)

    UF6 cylinders containing reactor grade enriched uranium are transported in protective overpacks. Recently, the design of the 2 1/2 ton UF6 cylinder overpack was modified to insure the safety of the cylinder inside the overpack. Modifications include a continuous stainless steel liner from the outer surface to the inner surface of the overpack and step joints between the upper and lower halves of the overpack. The effects of a continuous stainless steel liner and moisture in the insulation layer of a UF6 cylinder protective overpack were investigated with a numerical code. Results were compared with limited available field data. The purpose of comparing the numerical results with field data is to insure the validity of the numerical analysis and the physical properties used in the analysis. The study indicates that the continuous stainless steel liner did not influence the heat transfer rate much from the outer surface of the overpack to the 30B cylinder inside. The effect of step joints was not modeled due to the difficulty of quantifying the leakage rate through the gap. With a continuous stainless steel liner from the outside of the overpack to the inside, the overpack satisfies the thermal design criteria of protecting the cylinder inside for a minimum of 30 minutes when the overpack is exposed to a fire. The effect of moisture inside the insulation layer in the overpack is to reduce the energy to the cylinder with its high thermal capacity. The high pressure steam generated from the moisture will be relieved externally through the vent holes on the outer surface of the overpack. Although these holes are sealed after the overpack is dried, the plug sealing the holes will melt when the overpack is exposed to a fire

  7. Radiological considerations regarding an alternate method for the placement of intermediate impact absorbers at the Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    This report documents radiological considerations arising from the proposed implementation of an alternate method for intermediate impact absorber placement at the Canister Storage Building (CSB). These considerations include revising the dose rate estimate, at deck level over an open storage tube and outlining the administrative controls necessary for this implementation. Currently, the MCO Handling Machine (MHM) is used to install the intermediate impact absorbers. The proposed alternative would utilize a mobile crane, thus freeing up the MHM to handle the movement of MCOs within the CSB

  8. Trial manufacturing of titanium carbon-steel composite overpack for high-level radioactive waste disposal

    International Nuclear Information System (INIS)

    The overpack, a barrier enclosing the high-level radioactive waste (HLW), is designed to maintain complete containment for at least 1000 years. The titanium carbon-steel composite overpack adopts an outside titanium layer as a corrosion resistance to protect the inside carbon-steel body. The corrosion-proof overpack design could decrease the thickness of the shell, which complies with efficient space utilization in the disposal vault. Trial fabrication of actual size titanium carbon-steel composite overpack has been carried out to demonstrate the engineering feasibility and to extract the subjects for future improvement. The major dimensions of the cylindrical shape overpack are 1777 mm long, 914 mm outer diameter and 440 mm diameter hole to contain the HLW. Although manufacturing processes had not been optimized yet, the engineering feasibility of the titanium carbon-steel composite overpack was demonstrated successfully. (author)

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

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. UF6 cylinder overpack phenolic foam drop testing

    International Nuclear Information System (INIS)

    The types and quantities of materials to be used in the fire resistant phenolic foam for UF-6 cylinder protective overpacks are given in USDOE Material and Equipment Specification SP-9. Some of the specified materials and material grades used to make the foam have been unavailable or difficult to obtain since the late 1970s. Subsequently, overpack fabricators have found it necessary to substitute other materials or grades. With the requirements of SP-9 still applicable, it was necessary to determine if any property or quality of the phenolic foam was affected by the use of substituted materials in containers used to protect radioactive substances. The purpose of this report is to compare the mechanical shock absorbing ability of phenolic foam made from reagent grade chemicals specified in SP-9 to that of foam made from substituted commercial grade chemicals. The testing reported here consisted of mechanical drop tests of overpack models using foams made from different grades of the same chemicals and at different temperatures. These tests were performed to compare the mechanical properties of the foams

  12. Transportation and disposal of low-and medium level waste using fiber reinforced concrete overpacks

    International Nuclear Information System (INIS)

    A multiple-year research effort by Cogema culminated in the development of a new process to immobilize nuclear waste in concrete overpacks reinforced with metal fibers. The fiber concrete overpacks satisfy all French safety requirements relating to waste immobilization and disposal, and have been certified by Andra, the national radioactive waste management agency. This presentation will cover the use of the fiber-reinforced concrete overpack for disposal and transportation, and will discuss their fabrication. (J.P.N.)

  13. K-Basins particulate water content, and behavior

    International Nuclear Information System (INIS)

    This analysis summarizes the state of knowledge of K-basins spent nuclear fuel oxide (film, particulate or sludge) and its chemically bound water in order to estimate the associated multi-canister overpack (MCO) water inventory and to describe particulate dehydration behavior. This information can be used to evaluate the thermal and chemical history of an MCO and its contents during cold vacuum drying (CVD), shipping, and interim storage

  14. Advantages and limitations of navigation-based multicriteria optimization (MCO) for localized prostate cancer IMRT planning

    International Nuclear Information System (INIS)

    Efficacy of inverse planning is becoming increasingly important for advanced radiotherapy techniques. This study’s aims were to validate multicriteria optimization (MCO) in RayStation (v2.4, RaySearch Laboratories, Sweden) against standard intensity-modulated radiation therapy (IMRT) optimization in Oncentra (v4.1, Nucletron BV, the Netherlands) and characterize dose differences due to conversion of navigated MCO plans into deliverable multileaf collimator apertures. Step-and-shoot IMRT plans were created for 10 patients with localized prostate cancer using both standard optimization and MCO. Acceptable standard IMRT plans with minimal average rectal dose were chosen for comparison with deliverable MCO plans. The trade-off was, for the MCO plans, managed through a user interface that permits continuous navigation between fluence-based plans. Navigated MCO plans were made deliverable at incremental steps along a trajectory between maximal target homogeneity and maximal rectal sparing. Dosimetric differences between navigated and deliverable MCO plans were also quantified. MCO plans, chosen as acceptable under navigated and deliverable conditions resulted in similar rectal sparing compared with standard optimization (33.7 ± 1.8 Gy vs 35.5 ± 4.2 Gy, p = 0.117). The dose differences between navigated and deliverable MCO plans increased as higher priority was placed on rectal avoidance. If the best possible deliverable MCO was chosen, a significant reduction in rectal dose was observed in comparison with standard optimization (30.6 ± 1.4 Gy vs 35.5 ± 4.2 Gy, p = 0.047). Improvements were, however, to some extent, at the expense of less conformal dose distributions, which resulted in significantly higher doses to the bladder for 2 of the 3 tolerance levels. In conclusion, similar IMRT plans can be created for patients with prostate cancer using MCO compared with standard optimization. Limitations exist within MCO regarding conversion of navigated plans to

  15. Advantages and limitations of navigation-based multicriteria optimization (MCO) for localized prostate cancer IMRT planning

    Energy Technology Data Exchange (ETDEWEB)

    McGarry, Conor K., E-mail: conor.mcgarry@belfasttrust.hscni.net [Radiotherapy Physics, Northern Ireland Cancer Centre, Belfast Health and Social Care Trust, Belfast, Northern Ireland (United Kingdom); Bokrantz, Rasmus [Optimization and Systems Theory, KTH Royal Institute of Technology, Stockholm (Sweden); RaySearch Laboratories, Stockholm (Sweden); O’Sullivan, Joe M. [Centre for Cancer Research and Cell Biology, Queen’s University Belfast, Belfast, Northern Ireland (United Kingdom); Clinical Oncology, Northern Ireland Cancer Centre, Belfast Health and Social Care Trust, Belfast, Northern Ireland (United Kingdom); Hounsell, Alan R. [Radiotherapy Physics, Northern Ireland Cancer Centre, Belfast Health and Social Care Trust, Belfast, Northern Ireland (United Kingdom); Centre for Cancer Research and Cell Biology, Queen’s University Belfast, Belfast, Northern Ireland (United Kingdom)

    2014-10-01

    Efficacy of inverse planning is becoming increasingly important for advanced radiotherapy techniques. This study’s aims were to validate multicriteria optimization (MCO) in RayStation (v2.4, RaySearch Laboratories, Sweden) against standard intensity-modulated radiation therapy (IMRT) optimization in Oncentra (v4.1, Nucletron BV, the Netherlands) and characterize dose differences due to conversion of navigated MCO plans into deliverable multileaf collimator apertures. Step-and-shoot IMRT plans were created for 10 patients with localized prostate cancer using both standard optimization and MCO. Acceptable standard IMRT plans with minimal average rectal dose were chosen for comparison with deliverable MCO plans. The trade-off was, for the MCO plans, managed through a user interface that permits continuous navigation between fluence-based plans. Navigated MCO plans were made deliverable at incremental steps along a trajectory between maximal target homogeneity and maximal rectal sparing. Dosimetric differences between navigated and deliverable MCO plans were also quantified. MCO plans, chosen as acceptable under navigated and deliverable conditions resulted in similar rectal sparing compared with standard optimization (33.7 ± 1.8 Gy vs 35.5 ± 4.2 Gy, p = 0.117). The dose differences between navigated and deliverable MCO plans increased as higher priority was placed on rectal avoidance. If the best possible deliverable MCO was chosen, a significant reduction in rectal dose was observed in comparison with standard optimization (30.6 ± 1.4 Gy vs 35.5 ± 4.2 Gy, p = 0.047). Improvements were, however, to some extent, at the expense of less conformal dose distributions, which resulted in significantly higher doses to the bladder for 2 of the 3 tolerance levels. In conclusion, similar IMRT plans can be created for patients with prostate cancer using MCO compared with standard optimization. Limitations exist within MCO regarding conversion of navigated plans to

  16. Criticality safety evaluation report for spent nuclear fuelprocessing and storage facilities

    Energy Technology Data Exchange (ETDEWEB)

    Schwinkendorf, K.N., Fluor Daniel Hanford

    1997-03-24

    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 either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the MCO/Cask, operations at the Cold Vacuum Drying Facility (CVDF), and storage in the Canister Storage Building (CSB). Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} @ 0.95 criticality safety limit.

  17. 49 CFR 178.356 - Specification 20PF phenolic-foam insulated, metal overpack.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Specification 20PF phenolic-foam insulated, metal overpack. 178.356 Section 178.356 Transportation Other Regulations Relating to Transportation PIPELINE AND... Specification 20PF phenolic-foam insulated, metal overpack....

  18. Safety analysis report on the ''Paducah Tiger'' protective overpack for 10-ton cylinders of uranium hexafluoride

    International Nuclear Information System (INIS)

    The ''Paducah Tiger'' is a protective overpack used in shipment of 10-ton cylinders of enriched UF6. The calculations and tests are described which made and which indicate that the overpack is in compliance with the type B packaging requirements of ERDA Manual Chapter 0529 and Title 10 Code Federal Regulations Part 71. (U.S.)

  19. OPG Western waste management facility resin overpacking project

    International Nuclear Information System (INIS)

    Liners containing radioactive resins are stored in in-ground containers. Over time, degradation of the liners has occurred and there is potential for eventual leakage. The liners require overpacking in more robust packages to allow for extended storage and final placement in the Deep Geologic Repository. This paper will discuss the equipment design for safe venting, weather protection, radiation shielding, and remote handling of the liners. Alternative considerations and reasoning for final equipment design will be addressed. It will present issues encountered and how they were overcome as well as the logistical overview of the project, including milestones and time tables. (author)

  20. Nuclear Storage Overpack Door Actuator and Alignment Apparatus

    International Nuclear Information System (INIS)

    The invention is a door actuator and alignment apparatus for opening and closing the 15,000-pound horizontally sliding door of a storage overpack. The door actuator includes a ball screw mounted horizontally on a rigid frame including a pair of door panel support rails. An electrically powered ball nut moves along the ball screw. The ball nut rotating device is attached to a carriage. The carriage attachment to the sliding door is horizontally pivoting. Additional alignment features include precision cam followers attached to the rails and rail guides attached to the carriage

  1. Canister compatibility with Carlsbad salt

    International Nuclear Information System (INIS)

    No significant reaction was found when candidate canister alloys were heated with salt from Carlsbad, New Mexico, for up to 5000 hours in sealed capsules and for up to 10,000 hours in unsealed capsules at temperatures (80 to 2250C) that bracket the maximum temperature calculated for reference Savannah River Plant (SRP) waste containers at 20-foot spacings in salt. Additional tests were made at 6000C in sealed capsules to characterize reactions that may occur between candidate canister alloys and any component of the salt that is released when decrepitation occurs. Under these extreme conditions there was no significant attack of Type 304L stainless steel. But, there was up to 20-mils attack of the low-carbon steel

  2. Canister Transfer Facility Criticality Calculations

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Monroe-Rammsy

    2000-10-13

    The objective of this calculation is to evaluate the criticality risk in the surface facility for design basis events (DBE) involving Department of Energy (DOE) Spent Nuclear Fuel (SNF) standardized canisters (Civilian Radioactive Waste Management System [CRWMS] Management and Operating Contractor [M&O] 2000a). Since some of the canisters will be stored in the surface facility before they are loaded in the waste package (WP), this calculation supports the demonstration of concept viability related to the Surface Facility environment. The scope of this calculation is limited to the consideration of three DOE SNF fuels, specifically Enrico Fermi SNF, Training Research Isotope General Atomic (TRIGA) SNF, and Mixed Oxide (MOX) Fast Flux Test Facility (FFTF) SNF.

  3. 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)

  4. Retrievability of spent nuclear fuel canisters

    International Nuclear Information System (INIS)

    As a part of the designing process of the Finnish spent nuclear fuel repository, a preliminary study has been carried out to investigate how the canisters could technically be retrieved to the ground surface. Possibility of retrieving a canister has been investigated in different phases of the disposal project. Retrievability has not been a design goal for the spent fuel repository. However, design of the repository includes some features that may ease the retrieval of canisters in the future. Spent fuel elements are packaged in massive copper-iron canisters, which are mechanically strong and long-lived. The repository consists of excavated tunnels in hard rock which are supposed to be very long-lived making the removal of the tunnel backfilling technically possible also in the future. As long as the bentonite buffer has not been installed the canister can be returned to the ground surface using the same equipment as was used when the canister was brought down to the repository and lowered into the hole. In the encapsulation station the spent fuel elements can be packaged in the other canister or in the transport cask. After a deposition tunnel has been backfilled and closed, the retrieval consists of tearing down the concrete structure at the entry of the deposition tunnel, removal of the tunnel backfilling, removal of the bentonite from the disposal hole and lifting up of the canister. Various methods, e.g., flushing the bentonite with saline solutions, can be used to detach the canister from a hole with fully saturated bentonite. Recovery will be technically possible also after closing of the disposal facility. Backfilling of the shafts and tunnels will be removed and additional new structures and systems will have to be built in the repository. After that canisters can be transported to the ground surface as described above. In addition, handling of the canisters at the ground surface will require additional facilities. Canisters can be packaged in the

  5. Shielded canister transporter equipment acceptance test operations

    International Nuclear Information System (INIS)

    The defense waste processing facility (DWPF) processes high level waste at the Savannah River Plant (SRP) by vitrifying the waste and placing it in stainless stell canisters for long term storage. The shielded canister transporter (SCT) is a diesel powered mobile rubber tired self-propelled vehicle which transports the canisters from the DWPF processing facility to the on-site waste storage building. The SCT has a system of automatic programmable logic controls (PLC) which provides operational handling control with a shielded transfer cask and associated canister positional equipment

  6. 'Scorpion-like' dithiocarbamato-carboxylate ligands for linking M(CO)3+ (M = Tc, Re)

    International Nuclear Information System (INIS)

    Complexation of M(CO)3+ (M = Tc, Re) with dithiocarbamato-carboxylate ligands was studied. It was found that the dithiocarbamate chelation unit has rather high affinity to M(CO)3+ species and the resulting complexes are extremely stable in a wide pH range. The complex M(CO)3[S2CN(CH2COO)2] is monomeric in solution (ESI MS) and dimeric in the crystal phase (single-crystal XRD). The 'scorpion-like' dithiocarbamato-carboxylate ligands were prepared; it was shown that C5 chain is long enough to reach back to the free coordination vacancy in M(CO)3+ core and to block it. (author)

  7. Canister Storage Building (CSB) safety analysis report, phase 3: Safety analysis documentation supporting CSB construction

    International Nuclear Information System (INIS)

    The US Department of Energy established the K Basins Spent Nuclear Fuel Project to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored under water in the Hanford Site's K Basins, which are located near the Columbia River. Recommendations for a series of aggressive projects to construct and operate systems and facilities to manage the safe removal of K Basins fuel were made in WHC-EP-0830, Hanford Spent Nuclear Fuel Recommended Path Forward, and its subsequent update, WHC-SD-SNF-SP-005, Hanford Spent Nuclear Fuel Project Integrated Process Strategy for K Basins Fuel. The integrated process strategy recommendations include the following steps: Fuel preparation activities at the K Basins, including removing the fuel elements from their K Basin canisters, separating fuel particulate from fuel elements and fuel fragments greater than 0.6 cm (0.25 in.) in any dimension, removing excess sludge from the fuel and fuel fragments by means of flushing, as necessary, and packaging the fuel into multicanister overpacks (MCOs); Removal of free water by draining and vacuum drying at a cold vacuum drying facility ES-122; Dry shipment of fuel from the Cold Vacuum Drying to the Canister Storage Building (CSB), a new facility in the 200 East Area of the Hanford Site

  8. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    This report presents the inspection techniques of disposal canister components. Manufacturing methods and a description of the defects related to different manufacturing methods are described briefly. The defect types form a basis for the design of non-destructive testing because the defect types, which occur in the inspected components, affect to choice of inspection methods. The canister components are to nodular cast iron insert, steel lid, lid screw, metal gasket, copper tube with integrated or separate bottom, and copper lid. The inspection of copper material is challenging due to the anisotropic properties of the material and local changes in the grain size of the copper material. The cast iron insert has some acoustical material property variation (attenuation, velocity changes, scattering properties), which make the ultrasonic inspection demanding from calibration point of view. Mainly three different methods are used for inspection. Ultrasonic testing technique is used for inspection of volume, eddy current technique, for copper components only, and visual testing technique are used for inspection of the surface and near surface area

  9. Creep properties of EB welded copper overpack at 125-175 deg C

    International Nuclear Information System (INIS)

    Electron Beam welds (EBW) chosen as primary sealing method by Posiva welding the over-pack canister lids of oxygen-free phosphorus micro-alloyed copper (OFP) have been tested for material properties relevant to long term creep life prediction. Creep rupture results are presented for the ruptured 175 deg C tests and for the ongoing long term tests at 150 deg C and 125 deg C. The current status (test time, creep strain and strain rate) of the ongoing tests are reported. The initial (175 deg C) results indicate that the EB welds are weaker than the parent material and that both round bar and spark eroded square test specimens produce weld strengths of about 0.75 at tests durations of 5000 h. The downward trend is however expected to continue for the longer test durations. The creep ductility shows decrease for the longer tests. Life estimates for the EB weld have been calculated at 100 deg C for both 50 and 80 MPa with the so far lowest measured EB weld strength factor (WSF=0.77). The state-of-the-art model on the available data give estimated lives of 21000 and 3000 years correspondingly. However, simulated to the expected temperature profile of the repository service the life fraction reached after 10000 years of service is 1 % and 7 % for the same stress levels. It is though important to remembered that the 80 MPa assumption is very conservative in nature and that the predictions do not take into account relaxation of stresses, further decline of the WSF or anisotropy of the weld and are therefore still to be considered indicative only. It is also to be remembered that there is only limited data in the long term regime for the weldments and that the estimates are based on the few EB data available in the public domain added with the Posiva data of this project. Improvement of the models and predictions are expected from the ongoing 125 deg C and 150 deg C long term tests. (orig.)

  10. 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.

  11. Thermal analysis of cold vacuum drying of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piepho, M.G.

    1998-07-20

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

  12. Thermal analysis of cold vacuum drying of spent nuclear fuel

    International Nuclear Information System (INIS)

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4)

  13. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    J.F. Beesley

    2005-04-21

    The purpose of this facility description document (FDD) is to establish requirements and associated bases that drive the design of the Canister Handling Facility (CHF), which will allow the design effort to proceed to license application. This FDD will be revised at strategic points as the design matures. This FDD identifies the requirements and describes the facility design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This FDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This FDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the facility. Knowledge of these requirements is essential in performing the design process. The FDD follows the design with regard to the description of the facility. The description provided in this FDD reflects the current results of the design process.

  14. Canister storage building natural phenomena design loads

    International Nuclear Information System (INIS)

    This document presents natural phenomena hazard (NPH) loads for use in the design and construction of the Canister Storage Building (CSB), which will be located in the 200 East Area of the Hanford Site

  15. Canister transfer into repository in shaft alternative

    International Nuclear Information System (INIS)

    In this report, a study of lift transportation of a massive canister for spent nuclear fuel is considered. The canister is transferred from ground level to repository, which lies in the depth of 400 to 500 m in the bedrock. The canister is a massive metal vessel, whose weight is 19 to 29 tons, and which is strongly irradiant (gamma and neutrons), and which contains 1.4 to 2.2 tons of very strongly radio-active material, the activity of the fuel should not be spread in the environment even during postulated accidents. The study observes that the lift alternative is possible to be built and through good design practices and good maintenance procedures its safety, reliability and usability can be kept on such high level that canister transport is estimated to be licensable. (orig.)

  16. Conceptual designs of radioactive canister transporters

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-02-01

    This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts.

  17. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

    This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts

  18. Statistical analysis of DWPF reference canister dimensions

    International Nuclear Information System (INIS)

    Twenty dimensional measurements were conducted on seven empty Defense Waste Processing Facility (DWPF) reference canisters. These measurements were repeated after the canisters were filled with simulated nuclear waste glass. An in-depth statistical analysis of the results indicated that changes do occur as a result of filling the steel canisters with glass poured at 1150 degree C for four of the parameters. While small, these changes were statistically significant. The analysis indicates the maximum dimensional change found to occur after the filling for each variable. Statistical tests were used to determine if canister dimensions do significantly change, and corresponding variance information is presented. The results showed that the four measured parameters affected by filling are bottom diameter, bottom end diameter flange tilt, and lower head mismatch. Significant variability also existed for height, upper weld, ID label, lower head mismatch, and lower head ovality due to the measurements coming from different canisters. Finally, lower head mismatch showed variability caused by the data being taken at different locations on the canister. This location effect did not affect any of the other variables in this way

  19. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

    To facilitate the evaluation of alternate canister designs, 2 canisters were outfitted with thermocouples at elevations of 1/2, 3 1/2, and 6 1/2 inches from the canister bottom. The canisters were fabricated from two inch diameter schedule 10 and two inch diameter schedule 40 stainless steel pipe. Each canister was filled with approximately 2 kilograms of 49 wt percent lanthanide (Ln) loaded 25SrABS glass during 5 inch Cylindrical Induction Melter (CIM5) runs for TTR Tasks 3.03 and 4.03. Melter temperature, total mass of glass poured, and the glass pour rates were almost identical in both runs. The schedule 40 canister has a slightly smaller ID compared to the schedule 10 canister and therefore filled to a level of 9.5 inches compared to 8.0 inches for the schedule 40 canister. The schedule 40 canister had an empty mass of 1906 grams compared to 919 grams for the schedule 10 canister. The schedule 10 canister was found to have a higher maximum surface temperature by about 50--100 C (depending on height) during the glass pour compared to the schedule 40 canister. The additional thermal mass of the schedule 40 canister accounts for this difference. Once filled with glass, each of the canisters cooled at about the same rate, taking about an hour to cool below a maximum surface temperature of 200 C. No significant deformation of the either of the canisters was visually observed

  20. Proton radioactivity: the case for 53mCo proton-emitter isomer

    International Nuclear Information System (INIS)

    The partial proton emission half-life for 53mCo unstable isomer is re-examined in the framework of a semiempirical model based on tunneling through a Coulomb-plus-centrifugal-plus-overlapping potential barrier within the spherical nucleus approximation. It is shown that the known measured half-life value of 17 s is compatible with a large prolate shape for 53mCo proton emitter and a high angular momentum l = 11 assigned to the proton transition to the ground-state of 52Fe. (author)

  1. Hydrogen combustion in an MCO during interim storage (fauske and associates report 99-14)

    Energy Technology Data Exchange (ETDEWEB)

    PLYS, M.G.

    1999-05-12

    Flammable conditions are not expected to develop in an MCO during interim storage. This report considers potential phenomena which, although not expected t o occur, could lead t o flammable conditions. For example, reactions of hydrogen w i t h fuel over decades a r e postulated t o lead t o flammable atmospheric mixtures. For the extreme cases considered in this report, the highest attainable post-combustion pressure is about 13 atmospheres absolute, almost a factor of two and a half below the MCO design pressure of 31 atmospheres.

  2. 40 CFR 265.316 - Disposal of small containers of hazardous waste in overpacked drums (lab packs).

    Science.gov (United States)

    2010-07-01

    ... OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Landfills § 265.316 Disposal of small containers of hazardous waste in overpacked drums (lab packs). Small containers of hazardous waste... hazardous waste in overpacked drums (lab packs). 265.316 Section 265.316 Protection of...

  3. Development of an overpack for the storage of high-level waste in Swiss granitic bedrock

    International Nuclear Information System (INIS)

    Current programs aimed at demonstrating the feasibility of safe final disposal of high-level nuclear waste in Switzerland envisage a repository in the crystalline bedrock of the north of the country. The groundwater is reducing, with a mineralization of typically 10 g·L-1. The corrosion studies carried out in Switzerland have shown that unalloyed steel is a suitable overpack material under the conditions expected in the repository. The necessary corrosion allowance for a lifetime of 1000 years is 50 mm. Design work, based on the use of a typical cast steel with a tensile strength of 400 MN·m-2, has led to a reference overpack concept for a disposal of vitrified HLW. This reference overpack is designed as a self-shielding, self-supporting, cylindrical shell with hemispherical ends. 16 refs

  4. Design review report for the Hanford K East and K West Basins MCO loading system

    International Nuclear Information System (INIS)

    This design report presents the final design of the MCO Loading System. The report includes final design drawings, a system description, failure modes and recovery plans, a system operational description, and stress analysis. Design comments from the final design review have been incorporated

  5. Complete misloading of a mark IV MCO with mark 1A fuel

    International Nuclear Information System (INIS)

    The purpose of this analysis is to determine the worst case scenario for a total misload of a Mark IV MCO with Mark 1A fuel and scrap. This study is not intended to classify any of the components of the baskets

  6. 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.)

  7. Remote controlled mover for disposal canister transfer

    International Nuclear Information System (INIS)

    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.)

  8. Probabilistic sensitivity analysis for the 'initial defect in the canister' reference model

    International Nuclear Information System (INIS)

    In Posiva Oy's Safety Case 'TURVA-2012' the repository system scenarios leading to radionuclide releases have been identified in Formulation of Radionuclide Release Scenarios. Three potential causes of canister failure and radionuclide release are considered: (i) the presence of an initial defect in the copper shell of one canister that penetrates the shell completely, (ii) corrosion of the copper overpack, that occurs more rapidly if buffer density is reduced, e.g. by erosion, (iii) shear movement on fractures intersecting the deposition hole. All three failure modes are analysed deterministically in Assessment of Radionuclide Release Scenarios, and for the 'initial defect in the canister' reference model a probabilistic sensitivity analysis (PSA) has been carried out. The main steps of the PSA have been: quantification of the uncertainties in the model input parameters through the creation of probability density distributions (PDFs), Monte Carlo simulations of the evolution of the system up to 106 years using parameters values sampled from the previous PDFs. Monte Carlo simulations with 10,000 individual calculations (realisations) have been used in the PSA, quantification of the uncertainty in the model outputs due to uncertainty in the input parameters (uncertainty analysis), and identification of the parameters whose uncertainty have the greatest effect on the uncertainty in the model outputs (sensitivity analysis) Since the biosphere is not included in the Monte Carlo simulations of the system, the model outputs studied are not doses, but total and radionuclide-specific normalised release rates from the near-field and to the biosphere. These outputs are calculated dividing the activity release rates by the constraints on the activity fluxes to the environment set out by the Finnish regulator. Two different cases are analysed in the PSA: (i) the 'hole forever' case, in which the small hole through the copper overpack remains unchanged during the assessment

  9. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

  10. TtMCO: A highly thermostable laccase-like multicopper oxidase from the thermophilic Thermobaculum terrenum

    DEFF Research Database (Denmark)

    Brander, Søren; Mikkelsen, Jørn Dalgaard; Kepp, Kasper Planeta

    2015-01-01

    hyperthermal habitat of the host. TtMCO was screened for activity against 56 chemically diverse substrates. It displayed limited activity on classical LMCO substrates, such as e.g. phenolics, transition metals, or bilirubin. Highest activities were observed for nitrogen-containing aromatic compounds, i.e. 1......This paper reports the identification, heterologous expression in Escherichia coli and characterization of TtMCO from the thermophilic bacterium Thermobaculum terrenum, the first laccase-like multi-copper oxidase (LMCO) from the distinct Phylum Chloroflexi. TtMCO has only 39% identity to its...

  11. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    A near-field performance evaluation of an Advanced Cold Process Canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally proposed by TVO. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions has been assumed to persist during the service life of the canister. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie localized corrosion for the steel or copper canisters can be dismissed as a failure mechanism. The evaluation of the effects of processes outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. This factor will ensure the safety of the concept. (orig.)

  12. Update and insulation testing for uranium hexafluoride transport overpacks. United States Department of Transportation specification 21PF-1

    International Nuclear Information System (INIS)

    The slightly enriched product UF6 shipped from the enriching plants for the world's nuclear power plants must be protected in order to conform to domestic and international transport regulations. The principal overpack currently in use is the US Department of Transportation (DOT) Specification 21PF-1 which protects Model 30 UF6 cylinders (Title 49, Code of Federal Regulations, Part 178.121, Specification 21PF-1; Fire and Shock Resistant, Phenolic-Foam Insulated, Metal Overpack. Specification 21PF-1 (Horizontal Loading Overpack)). Operational problems have developed from both design and lack of maintenance, resulting in the entry of water into the insulation zone. In order to minimize this water entry, design modifications are necessary to the 21PF-1 overpacks. Proposed modifications for existing overpacks are to be made only after any water absorbed within the phenolic foam insulation is reduced to an acceptable level. New 21PF-1 overpacks will be fabricated under an enhanced design. In both cases, proposed quality assurance/control requirements in the fabrication, modification, use and maintenance of the overpacks are applicable to fabricators, modifiers, owners and users. Design changes are reviewed in Part I. The phenolic foam is the thermal barrier of the protective overpacks, which maintains the UF6 below its triple point in the event of exposure to elevated temperatures. Evaluation of the thermal qualities of the overpack required extensive analytical modeling correlated with experimental measurement. An experimental programme was devised to measure the thermal conductivity and heat capacity of the phenolic foam from room temperature to approximately 1475 deg. F (1073K). The test programme, which consisted of the guarded hot plate method for thermal conductivity and drop calorimetry for heat capacity determination, is reviewed in Part II. (author)

  13. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 3000C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 3000C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 8000C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts

  14. Rehearsal: Sample Canister in Cleanroom (Animation)

    Science.gov (United States)

    2005-01-01

    [figure removed for brevity, see original site] Click on the image for Rehearsal: Sample Canister in Cleanroom animation This movie shows rehearsal of the initial processing of the sample return capsule when it is taken to a temporary cleanroom at Utah's Test and Training Range.

  15. Techniques for freeing deposited canisters. Final report

    International Nuclear Information System (INIS)

    Four different techniques for removal of the bentonite buffer around a deposited canister have been identified, studied and evaluated: mechanical, hydrodynamical, thermal, and electrical techniques. Different techniques to determine the position of the canister in the buffer have also been studied: mechanical, electromagnetic, thermal and acoustic techniques. The mechanical techniques studied are full-face boring, milling and core-drilling. It is expected that the bentonite can be machined relatively easily. It is assessed that cooling by means of flushing water over the outer surfaces of the tools is not feasible in view of the tendency of bentonite to form a gel. The mechanical techniques are characterized by the potential of damaging the canister, a high degree of complexity, and high requirements of energy/power input. The generated byproduct is solid and cannot be removed by means of flushing. Removal is assessed to be simplest in conjunction with full-face boring and most difficult when coredrilling is applied. The hydrodynamical techniques comprise high-pressure hydrodynamic techniques, where pressures above and below 100 bar, and low pressure hydrodynamical techniques (< 10 bar) are separated. At pressures above 100 bar, a water jet with a diameter of approximately a millimetre cuts through the material. If desired, sand can be added to the jet. At pressures below 100 bar the jet has a diameter of one or a few centimetres. The liquid contains a few percent of salt, which is essential for the efficiency of the process. The flushing is important not only because it removes the modified bentonite but also because it frees previously unaffected bentonite and thereby makes it accessible to chemical modification. All of the hydrodynamical techniques are applicable for freeing the end surface as well as the mantle surface. The degree of complexity and the requirement on energy/power decrease with a decrease in pressure. A significant potential for damaging the

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

    International Nuclear Information System (INIS)

    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

  17. 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.

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

    OpenAIRE

    Jones, Peter M.; George, Anthony M.

    2016-01-01

    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 ...

  19. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    The report provides a summary of the design of the canister for final disposal of 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 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  20. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    A near-field performance evaluation of an advanced cold process canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie local-iced corrosion for the steel or copper canisters can be dismissed as a failure mechanism; The evaluation of the effects of processed outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. For completeness also evaluation of post-failure behaviour was carried out. Analyses were focussed on low probability phenomena from faults in canisters. Some items were identified where further research is justified in order to increase knowledge of the phenomena and thus strengthen the confidence of safety margins. However, it can be concluded that the risks of these scenarios can be judged to be acceptable. This is due to the fact that firstly, the probability of occurrence of most of these scenarios can be controlled to a large extent through technical measures. Secondly, these analyses indicated that the consequences would not be severe

  1. 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

  2. Safety analysis report on the ''Paducah Tiger'' protective overpack for 10-ton cylinders of uranium hexafluoride. Supplement 1

    International Nuclear Information System (INIS)

    This supplement details design changes made to the ''Paducah Tiger'' since the issue date of the SAR, June 16, 1976. A 3/8-inch thick, 304L stainless steel plate has been added on the valve end of existing and future overpacks to provide increased puncture resistance and the overpack cavity has been modified to preclude incorrect loading of the type 48X cylinder. Temperature profiles of the ''Paducah Tiger'' during the 30-minute fire test are included

  3. Pressurization of whole element canister during staging

    International Nuclear Information System (INIS)

    An analytical model was developed to estimate the buildup of gas pressure for a single outer element in a hot cell test container for a post cold vacuum drying staging/storage test. This model considers various sources of gas generation and gas consumption as a function of time. In a canister containing spent nuclear fuel, hydrogen is generated from the reactions of uranium with free water or hydrated water, hydride decomposition, and radiolysis. The canister pressurization model predicts a stable pressure and a peak temperature during staging, with an assumption that a fuel element contains 40 gm of corrosion products and a decay heat of 2.07 or 1.06 Watts. Calculations were also performed on constant temperature tests for fuel elements containing varied amounts of sludge tested at 150, 125, 105, and 85 C. The pressurization model will be used to evaluate test results obtained from post-drying testing on whole fuel elements

  4. CANISTER HANDLING FACILITY - VENTILATION CONFINEMENT ZONING ANALYSIS

    International Nuclear Information System (INIS)

    The purpose of this calculation is to calculate the necessary airflow distribution used to size the HVAC equipment for the Canister Handling Facility. These results will be compared to the Heating and Cooling Load Calculation in detailed design. The calculations contained in this document were developed by DandE/Mechanical HVAC and are intended solely for the use of the DandE/Mechanical HVAC department in its work regarding the HVAC system for the Canister Handling Facility. Yucca Mountain Project personnel from the DandE/Mechanical HVAC department should be consulted before use of the calculations for purposes other than those stated herein or used by individuals other than authorized personnel in DandE/Mechanical HVAC department

  5. CANISTER HANDLING FACILITY - VENTILATION AIR CALCULATION

    International Nuclear Information System (INIS)

    The purpose of this analysis is to establish the preliminary Ventilation Confinement Zone for the Canister Handling Facility (CHF). The results of this document will be used to determine the air quantities for each VCZ that will eventually be reflected in the development of the Ventilation Flow Diagrams. The analyses contained in this document are developed by D and E/Mechanical HVAC and are intended solely for the use of the D and E/Mechanical HVAC in its work regarding Confinement Zoning Analysis for the Canister Handling Facility. Yucca Mountain Project personnel from D and E/Mechanical HVAC should be consulted before use of the analyses for purposes other than those stated herein or used by individuals other than authorized personnel in D and E/Mechanical HVAC

  6. OREGON STATE UNIVERSITY (OSU) TRAINING RESEARCH ISOTOPE GENERAL ATOMICS (TRIGA) OVERPACK CLOSURE WELDING PROCESS PARAMETER DEVELOPMENT and QUALIFICATION

    International Nuclear Information System (INIS)

    Spent Nuclear Fuel (SNF) from the Oregon State University (OSU) TRIGA(regsign) Reactor is currently being stored in thirteen 55-gallon drums at the Hanford Site's low-level burial grounds. This fuel is soon to be retrieved from buried storage and packaged into new containers (overpacks) for interim storage at the Hanford Interim Storage Area (ISA). One of the key activities associated with this effort is final closure of the overpack by welding. The OSU fuel is placed into an overpack, a head inserted into the overpack top, and welded closed. Weld quality, for typical welded fabrication, is established through post-weld testing and nondestructive examination (NDE); however, in this case, once the SNF is placed into the overpack, routine testing and NDE are not feasible. An alternate approach is to develop and qualify the welding process/parameters, demonstrate beforehand that they produce the desired weld quality, and then verify parameter compliance during production welding. Fluor engineers have developed a Gas Tungsten Arc Welding (GTAW) technique and parameters, demonstrating that weld quality requirements for closure of packaged SNF overpacks are met, using this alternate approach. The following reviews the activities performed for this development and qualification effort

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

    International Nuclear Information System (INIS)

    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

  8. COMSOL MULTIPHYSICS MODEL FOR DWPF CANISTER FILLING

    Energy Technology Data Exchange (ETDEWEB)

    Kesterson, M.

    2011-03-31

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

  9. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. A concept of a nonfissile uranium hexafluoride overpack for storage, transport, and processing of corroded cylinders

    International Nuclear Information System (INIS)

    There is a need to develop a means of safely transporting breached 48-in. cylinders containing depleted uranium hexafluoride (UF6) from current storage locations to locations where the contents can be safely removed. There is also a need to provide a method of safely and easily transporting degraded cylinders that no longer meet the US Department of Transportation (DOT) and American National Standards Institute, Inc., (ANSI) requirements for shipments of depleted UF6. A study has shown that an overpack can be designed and fabricated to satisfy these needs. The envisioned overpack will handle cylinder models 48G, 48X, and 48Y and will also comply with the ANSI N14.1 and the American Society of Mechanical Engineers (ASME) Sect. 8 requirements

  12. Safety analysis report on the ''Paducah Tiger'' overpack for 10-ton cylinder of uranium hexafluoride

    International Nuclear Information System (INIS)

    A summary of analysis performed to assess the puncture resistance of the Paducah Tiger under a particularly severe (worst case) orientation of the external puncture pin is presented. The six-inch diameter cylindrical puncture pin has been oriented to place its impact location immediately opposite the valve body mounted to the dished head of the uranium hexafluoride cylinder. The valve body is assumed to have a one-inch clearance relative to the inner wall of the overpack. Analysis indicates that significant residual kinetic energy remains in the system at the instant of overpack inner wall contact with the valve body. Thus, there is strong evidence suggesting that the valve body can be damaged, or sheared from the dished head of the UF6, under the assumed worst case impact orientation

  13. PROPERTIES OF FIBERBOARD OVERPACK MATERIAL IN THE 9975 SHIPPING PACKAGE FOLLOWING THERMAL AGING

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W

    2007-01-10

    Many radioactive material shipping packages incorporate cane fiberboard overpacks for thermal insulation and impact resistance. Mechanical, thermal and physical properties have been measured on cane fiberboard following thermal aging in several temperature/humidity environments. Several of the measured properties change significantly over time in the more severe environments, while other properties are relatively constant. These properties continue to be tracked, with the goal of developing a model for predicting a service life under long-term storage conditions.

  14. Mechanical behaviour of high level nuclear waste overpacks under repository loading and during welding

    International Nuclear Information System (INIS)

    One of the concepts for final disposal of high level nuclear waste in Switzerland consists of a mined repository approximately 1200 m deep in the crystalline bedrock of Northern Switzerland. In order to delay the return of the radionuclides to the biosphere, and to reduce their concentration there to acceptable levels, reliance is placed in the multiple safety barrier principle. In addition to the natural barriers the following engineered barriers are envisaged: the waste form itself (vitrified high level nuclear waste), an overpack, and a compacted bentonite backfill within which the overpack is placed horizontally in the axis of the repository gallery. The first part of the present work reports on the participation in the COMPAS project (COntainer Mechanical Performance ASsessment). This project was carried out within the framework of the European Atomic Energy Community's cost-sharing programme on 'Radioactive Waste Management and Disposal'. It was concerned with the mechanical performance of overpacks for vitrified high level nuclear waste. The second part of this work deals with the issue of stress corrosion cracking of the high level nuclear waste overpack from NAGRA which is made out of GS-40 cast steel. After a description of the material properties of GS-40 cast steel, the one-dimensional FIBRE model is discussed, which should give an insight into the parameters involved in a thermomechanical calculation of a welding process. The calculations were performed with the commercially available finite element code ADINA. The output from ADINA was used as input to the postprocessor ORVIRT when fracture calculations are considered. (author) figs., tabs., refs

  15. Effect of magnetite as a corrosion product on the corrosion of carbon steel overpack

    International Nuclear Information System (INIS)

    It is necessary to clear the effects of corrosion products on the corrosion life time of carbon steel overpack for geological isolation of high-level radioactive waste (HLW). Especially, it is important to understand the effects of magnetite because magnetite as a simulated corrosion product is reported to accelerate the corrosion rate of carbon steel. In this study, corrosion tests to reproduce the acceleration of corrosion due to magnetite was performed and the mechanism of the acceleration was investigated to evaluate the effects of magnetite as a corrosion product. Based on the results of experiments, following conclusions are obtained; (1) Magnetite powder accelerates the corrosion rate of carbon steel. The main reaction of corrosion under the presence of magnetite is the reduction of Fe(III) in magnetite to Fe(II), but the reaction of hydrogen generation is also accelerated. The contribution of hydrogen generation reaction was estimated to be about 30% in the total corrosion reaction based on the experimental result of immersion test under the presence of magnetite. (2) Actual corrosion products containing magnetite generated by the corrosion of carbon steel protect the metal from the propagation of corrosion. The corrosion depth of carbon steel overpack due to magnetite was estimated to be about 1 mm based on the results of experiments. Even if the effect of magnetite is taken into the assessment of corrosion lifetime of overpack, total corrosion depth in 1000 years is estimated to be 33 mm, which is smaller than the corrosion allowance of 40 mm described in the second progress report on research and development for the geological disposal of HLW in Japan. It was concluded that the effect of magnetite on the corrosion life time of carbon steel overpack is negligible. (author)

  16. Testing in support of on-site storage of residues in the Pipe Overpack Container

    International Nuclear Information System (INIS)

    The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plans call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. It is desirable to use this same waste packaging for interim on-site storage in non-hardened buildings. To meet the safety concerns for this storage the Pipe Overpack Container has been subjected to a series of tests at Sandia National Laboratories in Albuquerque, New Mexico. In addition to the tests required to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II several tests were performed solely for the purpose of qualifying the container for interim storage. This report will describe these tests and the packages response to the tests. 12 figs., 3 tabs

  17. 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

  18. Stress corrosion cracking of copper canisters

    International Nuclear Information System (INIS)

    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

  19. Cold vacuum drying residual free water test description

    International Nuclear Information System (INIS)

    Residual free water expected to remain in a Multi-Canister Overpack (MCO) after processing in the Cold Vacuum Drying (CVD) Facility is investigated based on three alternative models of fuel crevices. Tests and operating conditions for the CVD process are defined based on the analysis of these models. The models consider water pockets constrained by cladding defects, water constrained in a pore or crack by flow through a porous bed, and water constrained in pores by diffusion. An analysis of comparative reaction rate constraints is also presented indicating that a pressure rise test can be used to show MCO's will be thermally stable at operating temperatures up to 75 C

  20. Hot isostatic pressing of copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    This paper describes the copper canisters designed by the Swedes for nuclear waste disposal. The canister is a large, plain, cylindrical can into which the spent nuclear fuel elements can be packed and sealed for final disposal. Two canister modifications are shown which have been developed, differing only in the method of packing the fuel elements into the canister. Both design approaches use a heavy-wall copper tube as the main body with forged end pieces machined to fit snugly on the tube. The favored approach today is the use of copper powder to surround the fuel elements, rather than lead. The canisters described were inserted into the chamber of a hot isostatic press machine. The result of subjecting the evacuated canister assembly to the combination of high temperature and pressure is compaction and densification of the entire mass and the conversion of the copper powder into a solid mass of copper. As a result of the hot isostatic pressing, the overall volume of the canister is reduced and the canister takes on a very moderate hourglass shape. These prototype canisters are sectioned and examined. The examination confirms that the process worked and that the result was of high quality

  1. Shaft shock absorber for a spent fuel canister

    International Nuclear Information System (INIS)

    The disposal canister for spent nuclear fuel will be transferred by a lift to the repository which is 500 m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorber can be developed against the design accident case where the canister should survive a free fall to the lift shaft. If the velocity of the canister is not controlled by air drag or by any other deceleration means, the impact velocity may reach ultimate speed of 100 m/s. The canister would retain its integrity in impact on water when the bottom pit of the lift well is filled with groundwater. However, the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20 m high filling to the bottom pit of the lift well by Light Expanded Clay Aggregate (LECA), gives fair impact absorption to protect the fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (author)

  2. 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.

  3. Cold Vacuum Drying (CVD) Facility Vacuum Purge System Chilled Water System Design Description (SYS 47-4)

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    2000-06-13

    This system design description (SDD) addresses the Vacuum Purge System Chilled Water (VPSCHW) system. The discussion that follows is limited to the VPSCHW system and its interfaces with associated systems. The reader's attention is directed to Drawings H-1-82162, Cold Vacuum Drying Facility Process Equipment Skid P&ID Vacuum System, and H-1-82224, Cold Vacuum Drying Facility Mechanical Utilities Process Chilled Water P&ID. Figure 1-1 shows the location and equipment arrangement for the VPSCHW system. The VPSCHW system provides chilled water to the Vacuum Purge System (VPS). The chilled water provides the ability to condense water from the multi-canister overpack (MCO) outlet gases during the MCO vacuum and purge cycles. By condensing water from the MCO purge gas, the VPS can assist in drying the contents of the MCO.

  4. Pitting corrosion on a copper canister

    International Nuclear Information System (INIS)

    It is demonstrated that normal pitting can occur during oxidizing conditions in the repository. It is also concluded that a new theory for pitting corrosion has to be developed, as the present theory is not in accordance with all practical and experimental observations. A special variant of pitting, based on the growth of sulfide whiskers, is suggested to occur during reducing conditions. However, such a mechanism needs to be demonstrated experimentally. A simple calculational model of canister corrosion was developed based on the results of this study. 69 refs, 3 figs

  5. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

    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''

  6. 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.

  7. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

    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

  8. Canister storage building trade study. Final report

    International Nuclear Information System (INIS)

    This study was performed to evaluate the impact of several technical issues related to the usage of the Canister Storage Building (CSB) to safely stage and store N-Reactor spent fuel currently located at K-Basin 100KW and 100KE. Each technical issue formed the basis for an individual trade study used to develop the ROM cost and schedule estimates. The study used concept 2D from the Fluor prepared ''Staging and Storage Facility (SSF) Feasibility Report'' as the basis for development of the individual trade studies

  9. Choices of canisters and elements for the first fuel and canister sludge shipment from K East Basin

    International Nuclear Information System (INIS)

    The K East Basin contains open-top canisters with up to fourteen N Reactor fuel assemblies distributed between the two barrels of each canister. Each fuel assembly generally consists of inner and outer concentric elements fabricated from uranium metal with zirconium alloy cladding. The canisters also contain varying amounts of accumulated sludge. Retrieval of sample fuel elements and associated sludge for examination is scheduled to occur in the near future. The purpose of this document is to specify particular canisters and elements of interest as candidate sources of fuel and sludge to be shipped to laboratories

  10. 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

  11. Impact testing of simulated high-level waste glass canisters

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, M.E.; Alzheimer, J.M.; Slate, S.C.

    1985-01-01

    Three Savannah River Laboratory reference high-level waste canisters were subjected to impact tests at the Pacific Northwest Laboratory in Richland, Washington, in June 1983. The purpose of the test was to determine the integrity of the canister, nozzle, and final closure weld and to assess the effects of impacts on the glass. Two of the canisters were fabricated from 304L stainless steel and the third canister from titanium. The titanium canister was subjected to two drops. The first drop was vertical from 9.14 m onto an unyielding surface with the bottom corner of the canister receiving the impact. No failure occurred during this drop. The second drop was vertical from 9.14 m onto an unyielding surface with the corner of the fill nozzle receiving the impact. A large breach in the canister occurred in the region where the fill nozzle joins the dished head. The first stainless steel canister was dropped with the corner of the fill nozzle receiving the impact. The canister showed significant strain with no rupturing in the region where the fill nozzle joins the dished head. The second canister was dropped with the bottom corner receiving the impact and also, dropped horizontally onto an unyielding vertical solid steel cylinder in a puncture test. The bottom drop did not damage the weld and the puncture test did not rupture the canister body. The glass particles in the damaged zone of these canisters were sampled and analyzed for particle size. A comparison was made with control canister in which no impact had occurred. The particle size distribution for the control canisters and the zones of damaged glass were determined down to 1.5 ..mu..m. The quantity of glass fines, smaller than 10 ..mu..m, which must be determined for transportation safety studies, was found to be the largest in the bottom-damaged zone. The total amount of fines smaller than 10 ..mu..m after impact was less than 0.01 wt % of the total amount of glass in the canister.

  12. Design analysis report for the canister

    International Nuclear Information System (INIS)

    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

  13. West Valley Demonstration Project full-scale canister impact tests

    International Nuclear Information System (INIS)

    Five West Valley Nuclear Services (WVNS) high-level waste (HLW) canisters were impact tested during 1994 to demonstrate compliance with the drop test requirements of the Waste Acceptance Product Specifications. The specifications state that the canistered waste form must be able to survive a 7-m (23 ft) drop unbreached. The 10-gauge (0.125 in. wall thickness) stainless steel canisters were approximately 85% filled with simulated vitrified waste and weighed about 2100 kg (4600 lb). Each canister was dropped vertically from a height of 7 m (23 ft) onto an essentially unyielding surface. The integrity of the canister was determined by the application and analysis of strain circles, dimensional measurements, and helium leak testing. The canisters were also visually inspected before and after the drop for physical damage. The results of the impact test verify that the canisters survived the 7-m drops unbreached. Therefore, these results demonstrate that the reference canister meets the drop test specification of the Waste Acceptance Product Specification

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

    International Nuclear Information System (INIS)

    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. Design basis for the copper/steel canister

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-02-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the point of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress which have been made between March 1995 and Feb 1996 and the result of further literature studies. A first trial canister has been produced using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. Similar problems exist with plate used for the fabricated tubular, but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. Welding of lids and bottoms to the copper canister is problematical.There is as yet no satisfactory non destructive test procedures for the parent metal or the welds in the copper canister material, partly due to the coarse grain size which arise in the proposed material processed by the proposed routes. Further studies are also required on crevice corrosion, galvanic attack and stress corrosion cracking in the copper 50 ppm phosphorus alloy. 28 refs.

  16. 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.

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

    International Nuclear Information System (INIS)

    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

  18. Performance assessment of the copper/steel canister

    International Nuclear Information System (INIS)

    A limited performance assessment has been done for a new canister concept. The assessment focuses primarily on a few specific questions. The areas given specific attention are: scenario development methodology, the effect of corrosion products, hydrogen gas transport and the retardation effect of the canister internals

  19. Design basis for the copper/steel canister

    International Nuclear Information System (INIS)

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the point of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress which have been made between March 1995 and Feb 1996 and the result of further literature studies. A first trial canister has been produced using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. Similar problems exist with plate used for the fabricated tubular, but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. Welding of lids and bottoms to the copper canister is problematical.There is as yet no satisfactory non destructive test procedures for the parent metal or the welds in the copper canister material, partly due to the coarse grain size which arise in the proposed material processed by the proposed routes. Further studies are also required on crevice corrosion, galvanic attack and stress corrosion cracking in the copper 50 ppm phosphorus alloy. 28 refs

  20. BRIC-100VC Biological Research in Canisters (BRIC)-100VC

    Science.gov (United States)

    Richards, Stephanie E.; Levine, Howard G. (Compiler); Romero, Vergel

    2016-01-01

    The Biological Research in Canisters (BRIC) is an anodized-aluminum cylinder used to provide passive stowage for investigations of the effects of space flight on small specimens. The BRIC 100 mm petri dish vacuum containment unit (BRIC-100VC) has supported Dugesia japonica (flatworm) within spring under normal atmospheric conditions for 29 days in space and Hemerocallis lilioasphodelus L. (daylily) somatic embryo development within a 5% CO2 gaseous environment for 4.5 months in space. BRIC-100VC is a completely sealed, anodized-aluminum cylinder (Fig. 1) providing containment and structural support of the experimental specimens. The top and bottom lids of the canister include rapid disconnect valves for filling the canister with selected gases. These specialized valves allow for specific atmospheric containment within the canister, providing a gaseous environment defined by the investigator. Additionally, the top lid has been designed with a toggle latch and O-ring assembly allowing for prompt sealing and removal of the lid. The outside dimensions of the BRIC-100VC canisters are 16.0 cm (height) x 11.4 cm (outside diameter). The lower portion of the canister has been equipped with sufficient storage space for passive temperature and relative humidity data loggers. The BRIC- 100VC canister has been optimized to accommodate standard 100 mm laboratory petri dishes or 50 mL conical tubes. Depending on storage orientation, up to 6 or 9 canisters have been flown within an International Space Station (ISS) stowage locker.

  1. 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.

  2. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    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

  3. Over-packing of spent ion exchange resin carbon steel liners at Ontario Power Generation's Western Waste Management Facility

    International Nuclear Information System (INIS)

    Spent resins from Ontario Power Generation (OPG)'s and Bruce Power's CANDU reactor operations are stored at OPG's Western Waste Management Facility in Kincardine, Ontario. The older resins are contained in 3 m3 epoxy-coated cylindrical carbon steel containers known as resin liners. These liners are stored in a stacked configuration within cylindrical in-ground containers. Previous studies indicated evidence of unacceptable liner wall corrosion and the potential for eventual leakage of resin from the liners. Based on this, OPG elected to repackage the majority of the resin liners into stainless steel over-packs. A contract for this work was awarded in mid-2006 to a project team consisting of Duratek of Canada, Kinectrics, Inc. and E.S. Fox. The project is expected to be completed by March 2008. This paper presents a) a description of the technical basis for overpacking and b) an overall summary of the over-packing project activities and achievements. (author)

  4. Description of DWPF reference waste form and canister

    International Nuclear Information System (INIS)

    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 requireiajps. The rabarajca saspa bkri is bkrksilicapa class cojtaining 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% SiO2 and 15% B2O3. This composition results in a low average leachability in the waste form of approximately 5 x 10-9 g/cm2-day based on 137Cs over 365 days in 250C 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 104 rem/hour at 1 cm

  5. 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.

  6. Aging Model For Cane Fiberboard Overpack In The 9975 Shipping Package

    International Nuclear Information System (INIS)

    Many radioactive material shipping packages incorporate a cane fiberboard overpack for thermal insulation and impact resistance. Mechanical, thermal and physical properties have been measured on cane fiberboard following thermal aging in several temperature/humidity environments. Several of the measured properties change significantly over time in the more severe environments, while other properties are relatively constant. Changes in each of the properties have been fit to a model to allow predictions of degradation under various storage scenarios. Additional data continue to be collected to provide for future refinements to the model.

  7. 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

  8. 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

  9. Criticality safety calculations of storage canisters

    International Nuclear Information System (INIS)

    In the planned Swedish repository for deep disposal of spent nuclear fuel the fuel assemblies will be stored in storage canisters made of cast iron and copper. To assure safe storage of the fuel the requirement is that the normal criticality safety criteria have to be met. The effective neutron multiplication factor must not exceed 0.95 in the most reactive conditions including different kinds of uncertainties. In this report it is shown that the criteria could be met if credit for the reactivity decrease due to the burn up of the fuel is taken into account. The criticality safety criteria are based on the US NRC regulatory requirements for transportation and storage of spent fuel

  10. Canister Storage Building (CSB) Hazard Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    2000-03-16

    This report describes the methodology used in conducting the Canister Storage Building (CSB) Hazard Analysis to support the final CSB Safety Analysis Report and documents the results. 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 process identified hazardous conditions and material-at-risk, determined causes for potential accidents, identified preventive and mitigative features, and qualitatively estimated the frequencies and consequences of specific occurrences. The hazard analysis was performed by a team of cognizant CSB operations and design personnel, safety analysts familiar with the CSB, and technical experts in specialty areas. The material included in this report documents the final state of a nearly two-year long process. Attachment A provides two lists of hazard analysis team members and describes the background and experience of each. The first list is a complete list of the hazard analysis team members that have been involved over the two-year long process. The second list is a subset of the first list and consists of those hazard analysis team members that reviewed and agreed to the final hazard analysis documentation. The material included in this report documents the final state of a nearly two-year long process involving formal facilitated group sessions and independent hazard and accident analysis work. The hazard analysis process led to the selection of candidate accidents for further quantitative analysis. New information relative to the hazards, discovered during the accident analysis, was incorporated into the hazard analysis data in order to compile a complete profile of facility hazards. Through this process, the results of the hazard and accident analyses led directly to the identification of safety structures, systems, and components, technical safety requirements, and other

  11. Shippingport Spent Fuel Canister System Description

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-03-27

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available.

  12. Canister Storage Building (CSB) Hazard Analysis Report

    International Nuclear Information System (INIS)

    This report describes the methodology used in conducting the Canister Storage Building (CSB) Hazard Analysis to support the final CSB Safety Analysis Report and documents the results. 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 process identified hazardous conditions and material-at-risk, determined causes for potential accidents, identified preventive and mitigative features, and qualitatively estimated the frequencies and consequences of specific occurrences. The hazard analysis was performed by a team of cognizant CSB operations and design personnel, safety analysts familiar with the CSB, and technical experts in specialty areas. The material included in this report documents the final state of a nearly two-year long process. Attachment A provides two lists of hazard analysis team members and describes the background and experience of each. The first list is a complete list of the hazard analysis team members that have been involved over the two-year long process. The second list is a subset of the first list and consists of those hazard analysis team members that reviewed and agreed to the final hazard analysis documentation. The material included in this report documents the final state of a nearly two-year long process involving formal facilitated group sessions and independent hazard and accident analysis work. The hazard analysis process led to the selection of candidate accidents for further quantitative analysis. New information relative to the hazards, discovered during the accident analysis, was incorporated into the hazard analysis data in order to compile a complete profile of facility hazards. Through this process, the results of the hazard and accident analyses led directly to the identification of safety structures, systems, and components, technical safety requirements, and other

  13. Shippingport Spent Fuel Canister System Description

    International Nuclear Information System (INIS)

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available

  14. US NRC-Sponsored Research on Stress Corrosion Cracking Susceptibility of Dry Storage Canister Materials in Marine Environments - 13344

    International Nuclear Information System (INIS)

    At a number of locations in the U.S., spent nuclear fuel (SNF) is maintained at independent spent fuel storage installations (ISFSIs). These ISFSIs, which include operating and decommissioned reactor sites, Department of Energy facilities in Idaho, and others, are licensed by the U.S. Nuclear Regulatory Commission (NRC) under Title 10 of the Code of Federal Regulations, Part 72. The SNF is stored in dry cask storage systems, which most commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack vented to the external atmosphere to allow airflow for cooling. Some ISFSIs are located in marine environments where there may be high concentrations of airborne chloride salts. If salts were to deposit on the canisters via the external vents, a chloride-rich brine could form by deliquescence. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC), particularly in the presence of residual tensile stresses from welding or other fabrication processes. SCC could allow helium to leak out of a canister if the wall is breached or otherwise compromise its structural integrity. There is currently limited understanding of the conditions that will affect the SCC susceptibility of austenitic stainless steel exposed to marine salts. NRC previously conducted a scoping study of this phenomenon, reported in NUREG/CR-7030 in 2010. Given apparent conservatisms and limitations in this study, NRC has sponsored a follow-on research program to more systematically investigate various factors that may affect SCC including temperature, humidity, salt concentration, and stress level. The activities within this research program include: (1) measurement of relative humidity (RH) for deliquescence of sea salt, (2) SCC testing within the range of natural absolute humidity, (3) SCC testing at elevated temperatures, (4) SCC testing at high humidity conditions, and (5) SCC testing with various applied stresses. Results

  15. 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

  16. Testing in support of transportation of residues in the pipe overpack container

    International Nuclear Information System (INIS)

    The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plants call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. The tests described here were performed to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II. Using a more robust container will assure the fissile materials in each container can not be mixed with the fissile material from the other containers and will provide criticality control. This will allow an increase in the payload of the TRUPACT-II from 325 fissile gram equivalents to 2,800 fissile gram equivalents

  17. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    The holding canister for spent nuclear fuel will be transferred by a lift to the final disposal tunnels 500m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorbing properties can be met in a design accident case where the canister should survive a free fall due to e.g. sabotage. If the velocity of the canister is not controlled by air drag or any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity when stricken by the surface penetration impact if the bottom pit of the lift well would be filled with groundwater. However the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20m high filling to the bottom pit of the lift well by ceramic gravel, trade mark LECA-sora, gives a fair impact absorption to protect the spent fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

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

    International Nuclear Information System (INIS)

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

  19. Evaluation of remote smearing of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    The Savannah River Site (SRS) is evaluating the variables of the remote smearing process for monitoring transferable contamination on the waste glass canisters at the Defense Waste Processing Facility (DWPF). Smearing for transferable contamination is typically done by hand, but in this case, due to the nature of the high level waste within the canisters, remote smearing is required. The effectiveness of the smear pad was determined under varying conditions (distance traveled, force applied, and canister surface), as well as the relative importance of these factors. It was concluded that the remote smear is more reliable than the hand smear

  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. Containment canister for capturing hazardous waste debris during piping modifications

    Science.gov (United States)

    Dozier, Stanley B.

    2001-07-24

    The present invention relates to a capture and containment canister which reduces the risk of radiation and other biohazard exposure to workers, the need for a costly containment hut and the need for the extra manpower associated with the hut. The present invention includes the design of a canister having a specially designed magnetic ring that attracts and holds the top of the canister in place during modifications to gloveboxes and other types of radiological and biochemical hoods. The present invention also provides an improved hole saw that eliminates the need for a pilot bit.

  2. Retrievability of spent nuclear fuel canisters; Kaeytetyn ydinpolttoaineen loppusijoituskapseleiden palautettavuus

    Energy Technology Data Exchange (ETDEWEB)

    Saanio, T. [Saanio and Riekkola Oy, Helsinki (Finland); Raiko, H. [VTT Energy, Espoo (Finland)

    1999-03-01

    As a part of the designing process of the Finnish spent nuclear fuel repository, a preliminary study has been carried out to investigate how the canisters could technically be retrieved to the ground surface. Possibility of retrieving a canister has been investigated in different phases of the disposal project. Retrievability has not been a design goal for the spent fuel repository. However, design of the repository includes some features that may ease the retrieval of canisters in the future. Spent fuel elements are packaged in massive copper-iron canisters, which are mechanically strong and long-lived. The repository consists of excavated tunnels in hard rock which are supposed to be very long-lived making the removal of the tunnel backfilling technically possible also in the future. As long as the bentonite buffer has not been installed the canister can be returned to the ground surface using the same equipment as was used when the canister was brought down to the repository and lowered into the hole. In the encapsulation station the spent fuel elements can be packaged in the other canister or in the transport cask. After a deposition tunnel has been backfilled and closed, the retrieval consists of tearing down the concrete structure at the entry of the deposition tunnel, removal of the tunnel backfilling, removal of the bentonite from the disposal hole and lifting up of the canister. Various methods, e.g., flushing the bentonite with saline solutions, can be used to detach the canister from a hole with fully saturated bentonite. Recovery will be technically possible also after closing of the disposal facility. Backfilling of the shafts and tunnels will be removed and additional new structures and systems will have to be built in the repository. After that canisters can be transported to the ground surface as described above. In addition, handling of the canisters at the ground surface will require additional facilities. Canisters can be packaged in the

  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. 49 CFR 178.358 - Specification 21PF fire and shock resistant, phenolic-foam insulated, metal overpack.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Specification 21PF fire and shock resistant, phenolic-foam insulated, metal overpack. 178.358 Section 178.358 Transportation Other Regulations Relating... Class 7 (Radioactive) Materials § 178.358 Specification 21PF fire and shock resistant,...

  5. 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.

  6. 40 CFR 264.316 - Disposal of small containers of hazardous waste in overpacked drums (lab packs).

    Science.gov (United States)

    2010-07-01

    ... HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Landfills § 264.316 Disposal of small containers... CFR parts 173, 178, and 179), if those regulations specify a particular inside container for the waste... hazardous waste in overpacked drums (lab packs). 264.316 Section 264.316 Protection of...

  7. Thermal dimensioning of the deep repository. Influence of canister spacing, canister power, rock thermal properties and nearfield design on the maximum canister surface temperature

    Energy Technology Data Exchange (ETDEWEB)

    Hoekmark, Harald; Faelth, Billy [Clay Technology AB, Lund (Sweden)

    2003-12-01

    The report addresses the problem of the minimum spacing required between neighbouring canisters in the deep repository. That spacing is calculated for a number of assumptions regarding the conditions that govern the temperature in the nearfield and at the surfaces of the canisters. The spacing criterion is that the temperature at the canister surfaces must not exceed 100 deg C .The results are given in the form of nomographic charts, such that it is in principle possible to determine the spacing as soon as site data, i.e. the initial undisturbed rock temperature and the host rock heat transport properties, are available. Results of canister spacing calculations are given for the KBS-3V concept as well as for the KBS-3H concept. A combination of numerical and analytical methods is used for the KBS-3H calculations, while the KBS-3V calculations are purely analytical. Both methods are described in detail. Open gaps are assigned equivalent heat conductivities, calculated such that the conduction across the gaps will include also the heat transferred by radiation. The equivalent heat conductivities are based on the emissivities of the different gap surfaces. For the canister copper surface, the emissivity is determined by back-calculation of temperatures measured in the Prototype experiment at Aespoe HRL. The size of the different gaps and the emissivity values are of great importance for the results and will be investigated further in the future.

  8. Technical note 4. Corrosion of copper canister

    International Nuclear Information System (INIS)

    Objectives of the project: In this review assignment, SKB's treatment of copper corrosion processes or mechanisms in SR-Site shall be reviewed both for the anticipated oxic and anoxic repository environments. The reviewer(s) shall consider if corrosion and corrosion mechanisms of the copper canisters in different possible evolutionary repository environments have been properly described. The objectives of this initial review phase in the area of copper corrosion is to achieve a broad coverage of SR-Site and its supporting references and in particular identify the need for complementary information and clarifications to be delivered by SKB. Summary by the authors: It is expected that the inflow of ground water to the deposition holes and tunnels in the Forsmark repository will be very slow. Thus, it might take some few hundred years up to thousand years before the deposition holes are filled with ground water and it might take 6000 years or more before the bentonite buffer is fully water saturated and pressurized. The copper canisters will therefore meet to two completely different environments: 1. An initial period of several hundreds of years when copper is exposed to gaseous corrosion. 2. And then to aqueous corrosion. From a corrosion point of view the first 1000 years are the most critical for the copper canister since pure, or phosphorus alloyed copper, is not designed to cope with corrosion at elevated temperatures. The outer copper surface temperature is expected to reach 100 deg C within some decades after closure of the repository and then slowly cool down to around 50 deg C after 1000 years. The gaseous corrosion is treated in SKB's safety assessment as being only dependent on oxygen gas and thus easily estimated by an oxygen mass-balance calculation. This simple model has no scientific support since several corrosive trace gases, such as sulphurous and nitrous compounds, operates together with water molecules (moisture) and the corrosion product consists

  9. Theoretical Basis for the Design of a DWPF Evacuated Canister

    Energy Technology Data Exchange (ETDEWEB)

    Routt, K.R.

    2001-09-17

    This report provides the theoretical bases for use of an evacuated canister for draining a glass melter. Design recommendations are also presented to ensure satisfactory performance in future tests of the concept.

  10. Canister design for deep borehole disposal of nuclear waste

    OpenAIRE

    Hoag, Christopher Ian.

    2006-01-01

    The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories using currently available and proven oil, gas, and geothermal drilling technology. The canister is suitable for disposal of various waste forms, such as fuel assemblies and vitrified waste. The design addresses real and perceived hazards of transporting and placing high-level waste, in the form of spent reactor fuel, into a deep igneous rock env...

  11. 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.

  12. Spent nuclear fuel canister storage building conceptual design report

    International Nuclear Information System (INIS)

    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

  13. 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.

  14. SLUDGE TREATMENT PROJECT KOP CONCEPTUAL DESIGN CONTROL DECISION REPORT

    International Nuclear Information System (INIS)

    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.

  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. Drop tests of the Three Mile Island knockout canister

    International Nuclear Information System (INIS)

    A type of Three Mile Island Unit 2 (TMI-2) defueling canister, called a ''knockout'' canister, was subjected to a series of drop tests at the Oak Ridge National Laboratory's Drop Test Facility. These tests were designed to confirm the structural integrity of internal fixed neutron poisons in support of a request for NRC licensing of this type of canister for the shipment of TMI-2 reactor fuel debris to the Idaho National Engineering Laboratory (INEL) for the Core Examination R and D Program. Work conducted at the Oak Ridge National Laboratory included (1) precise physical measurements of the internal poison rod configuration before assembly, (2) canister assembly and welding, (3) nondestructive examination (an initial hydrostatic pressure test and an x-ray profile of the internals before and after each drop test), (4) addition of a simulated fuel load, (5) instrumentation of the canister for each drop test, (6) fabrication of a cask simulation vessel with a developed and tested foam impact limiter, (7) use of refrigeration facilities to cool the canister to well below freezing prior to three of the drops, (8) recording the drop test with still, high-speed, and normal-speed photography, (9) recording the accelerometer measurements during impact, (10) disassembly and post-test examination with precise physical measurements, and (11) preparation of the final report

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

    International Nuclear Information System (INIS)

    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

  18. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

    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)

  19. 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.

  20. Corrosion evaluation of fuel canister crusher rigging

    International Nuclear Information System (INIS)

    A fuel canister crusher with attached rigging is located in the 105 K-East Basin discharge chute. This equipment is slated to be moved as part of seismic mitigation to prevent a major basin leak through a construction joint located in the base of the chute. This corrosion analysis assessed the load-bearing ability of the rigging, which consists of shackles and thimble-spliced wire rope. The K-East Basin demineralized water results in corrosion rates of <2 mil/year (<0.05 mm/year) for carbon, low-alloy carbon, and stainless steels. The galvanized carbon steel shackles (with low-alloy steel anchor pins) have experienced negligible corrosion and are judged to be mechanically unaffected by their water exposure. The carbon steel wire rope and stainless steel thimbles have undergone minimal corrosion. Due to the small amount of corrosion products (as seen from video inspection), the absence of wire breakage, and a Factor of Safety calculation, it is judged that the wire rope and thimbles would withstand the proposed relocation activities

  1. One-pot low temperature synthesis of MFe2O4 (M=Co, Ni, Zn) superparamagnetic nanocrystals

    International Nuclear Information System (INIS)

    Magnetic MFe2O4 (M=Co, Ni, Zn) nanocrystals with a diameter about 30 nm and a nearly spherical shape were synthesized via a simple hydrothermal approach. X-ray diffraction, scanning electron microscopy, and transmission electron microscopy have been used to investigate the as-prepared magnetic MFe2O4 (M=Co, Ni, Zn) nanocrystals. Magnetic properties of the as-prepared samples have been detected by a vibrating sample magnetometer at room temperature and the results show that the as-prepared magnetic MFe2O4 nanocrystals are a type characteristic of superparamagnetic materials. These superparamagnetic nanocrystals are believed to be promising for wide engineering applications, such as drug delivery, bioseparation, and magnetic resonance imaging

  2. Canister filling materials -- Design requirements and evaluation of candidate materials

    International Nuclear Information System (INIS)

    SKB has been evaluating a copper/steel canister for use in the disposal of spent nuclear reactor fuel. Once the canister is breached by corrosion, it is possible that the void volume inside the canister might fill with water. Water inside the canister would moderate the energy of the neutrons emitted by spontaneous fission in the fuel. It the space in the canister between and around the fuel pins is occupied by canister filling materials, the potential for criticality is avoided. The authors have developed a set of design requirements for canister filling material for the case where it is to be used alone, with no credit for burnup of the fuel or other measures, such as the use of neutron absorbers. Requirements were divided into three classes: essential requirements, desirable features, and undesirable features. The essential requirements are that the material fill at least 60% of the original void space, that the solubility of the filling material be less than 100 mg/l in pure water or expected repository waters at 50 C, and that the material not compact under its own weight by more than 10%. In this paper they review the reasons for these requirements, the desirable and undesirable features, and evaluate 11 candidate materials with respect to the design requirements and features. The candidate materials are glass beads, lead shot, copper spheres, sand, olivine, hematite, magnetite, crushed rock, bentonite, other clays, and concrete. Emphasis is placed on the determination of whether further work is needed to eliminate uncertainties in the evaluation of the ability of a particular filling material to be successfully used under actual conditions, and on the ability to predict the long-term performance of the material under the repository conditions

  3. Vitrification of high level wastes: a review of the computer thermal analyses for storage canisters

    International Nuclear Information System (INIS)

    CANIST, a two-dimensional (r and THETA) computer program that solves the unsteady-state, heat conduction equation was used to model the thermal behavior of canisters filled with waste glass. CANIST has been found to be a valuable analytical tool for predicting the temperature profile of a waste storage canister as a function of several variables, including the diameter of the canister, the placement of internal fins, the heat generation rate of the waste glass, and the thermophysical properties of the canister and the waste glass. Thus, temperature dependent processes that may affect the integrity of the glass/canister unit, for example cracking, can be investigated using an analytical approach. In the present study, the canister temperature profiles predicted by CANIST were compared to canister temperatures measured during full-scale non-radioactive waste immobilization tests conducted at Pacific Northwest Laboratory. The agreement between experimental and predicted temperatures was good, particularly considering the fact that the thermophysical properties of the waste glass modeled have not yet been accurately determined. Examination of some glass-filled canisters has revealed cracking to have occurred in the glass. However, the comparison between measured and CANIST predicted temperatures suggests that cracking does not significantly influence the heat-transfer process. CANIST was also used to evaluate different ways of reducing the centerline temperature of a canister, and to predict the centerline temperature as a function of the heat generation rate of the waste glass and the type of interim storage, i.e., air or water

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

    International Nuclear Information System (INIS)

    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. Mechanical Integrity of Canisters Using a Fracture Mechanics Approach

    International Nuclear Information System (INIS)

    This report presents the methods and results of a research project about numerical modeling of mechanical integrity of cast-iron canisters for the final disposal of spent nuclear fuel in Sweden, using combined boundary element (BEM) and finite element (FEM) methods. The objectives of the project are: 1) to investigate the possibility of initiation and growth of fractures in the cast-iron canisters under the mechanical loading conditions defined in the premises of canister design by Swedish Nuclear Fuel and Waste Management Co. (SKB); 2) to investigate the maximum bearing capacity of the cast iron canisters under uniformly distributed and gradually increasing boundary pressure until plastic failure. Achievement of the two objectives may provide some quantitative evidence for the mechanical integrity and overall safety of the cast-iron canisters that are needed for the final safety assessment of the geological repository of the radioactive waste repository in Sweden. The geometrical dimension, distribution and magnitudes of loads and Material properties of the canisters and possible fractures were provided by the latest investigations of SKB. The results of the BEM simulations, using the commercial code BEASY, indicate that under the currently defined loading conditions the possibility of initiation of new fractures or growth of existing fractures (defects) are very small, due to the reasons that: 1) the canisters are under mainly compressive stresses; 2) the induced tensile stress regions are too small in both dimension and magnitude to create new fractures or to induce growth of existing fractures, besides the fact that the toughness of the fractures in the cast iron canisters are much higher that the stress intensity factors in the fracture tips. The results of the FEM simulation show a approximately 75 MPa maximum pressure beyond which plastic collapse of the cast-iron canisters may occur, using an elastoplastic Material model. This figure is smaller compared

  6. The 200 l stainless steel canister - remote handling clutch assembly

    International Nuclear Information System (INIS)

    The assembly 200 l stainless steel canister with remote handling clutch is an equipment for conditioning, transport and intermediate storage of solid low- and intermediate level radioactive wastes. Loading the canister with pre-conditioned radioactive wastes is done at Post-Irradiation Examination Laboratory (LEPI) of INR Pitesti either within the transfer cell (CT) or supra-cell (SC). To this goal, lifting and handling means with which the LEPI is equipped, namely, lifting bridge and remote handling clutch are used. Conditioning of waste in view of their removal from LEPI implies their solidification in concrete and placing in stainless steel canister, the operations being effected in adequate rooms correspondingly equipped in the frame of the shop located at +8.40 m height at LEPI. Technical characteristics are: - capacity, 200 l; - external diameter, max. 600 mm; - casing height, 925 mm; casing thickness, 1.5 mm; - bottom thickness, 3 mm; - lid thickness, 3 mm. The canister cross profile of the lower and upper ends is modelled so that pilling is possible without horizontal slipping. The equipment together with remote handling clutch, engaged in a special collar of the upper part of canister, is presented

  7. 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

  8. Design, production and initial state of the canister

    International Nuclear Information System (INIS)

    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

  9. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  10. Proposal for modifications to US Department of Transportation specification 21PF-1 fire and shock resistant phenolic foam-insulated metal overpack

    International Nuclear Information System (INIS)

    Slightly enriched product uranium hexafluoride (UF6) shipped from enriching plants for the world's nuclear power plants must be protected in order to conform to domestic and international transport regulations. The principal overpack currently in use is the US Department of Transportation (DOT) Specification 21PF-1 which protects Model 30 UF6 cylinders (Title 49, Code of Federal Regulations, Part 178.121, Specification 21PF-1; Fire and Shock Resistant, Phenolic-Foam Insulated, Metal Overpack). Operational problems developed from both design and lack of maintenance have resulted in the entry of water into the insulation. To minimize this water entry, design modifications are necessary. Proposed modifications for existing overpacks are to be made only after any water absorbed within the insulation is reduced to an acceptable level. New 21PF-1 overpacks will be fabricated under an enhanced design. In both cases, proposed quality assurance/control requirements in the fabrication, modification, use and maintenance of the overpacks are applicable to fabricators, modifiers, owners and users. The phenolic foam is the thermal barrier, which maintains the UF6 below its triple point in the event of exposure to elevated temperatures. Evaluation of the thermal qualities of the overpack required extensive analytical modeling correlated with experimental measurement. An experimental program was devised to measure the thermal conductivity and heat capacity of the phenolic foam from room temperature to approximately 14750F (10730K)

  11. A modeling study of general corrosion of copper overpack for geological isolation of high-level radioactive waste

    International Nuclear Information System (INIS)

    This paper describes a modeling study for general corrosion of copper which is a candidate material for high-level radioactive waste overpacks. The model is a mixed-potential model combined with diffusive transport of reactants and reaction products. The rest potential and corrosion rate of copper in aerated solution were measured while controlling the thickness of a diffusive solution layer on the copper surface using a rotating-disk electrode. Experimental data were used for validation of the model

  12. Decontamination of DWPF canisters by glass frit blasting

    International Nuclear Information System (INIS)

    High-level radioactive waste at the Savannah River Plant will be incorporated in borosilicate glass for permanent disposal. The waste glass will be encapsulated in a 304L stainless steel canister. During the filling operation the outside of the canister will become contaminated. This contamination must be reduced to an accepable level before the canister leaves the Defense Waste Processing Facility (DWPF). Tests with contaminated coupons have demonstrated that this decontamination can be accomplished by blasting the surface with glass frit. The contaminated glass frit byproduct of this operation is used as a feedstock for the waste glass process, so no secondary waste is created. Three blasting techniques, using glass frit as the blasting medium, were evaluated. Air-injected slurry blasting was the most promising and was chosen for further development. The optimum parametric values for this process were determined in tests using coupon weight loss as the output parameter. 1 reference, 13 figures, 3 tables

  13. Radiation-field mapping of insect irradiation canisters

    International Nuclear Information System (INIS)

    Dosimetry methods developed at NIST for mapping ionizing radiation fields were applied to canisters used in 137Cs dry-source irradiators designed for insect sterilization. The method of mapping the radiation fields inside of these canisters as they cycled through the gamma-ray irradiators involved the use of radiochromic films, which increase in optical density proportionately to the absorbed dose. A dosimeter film array in a cardboard phantom was designed to simulate the average insect pupae density and to map the dose within the full volume of the canister; the calibrated films were read using a laser scanning densitometer. Previously used dosimetric methods did not allow for the spatial resolution that is possible with these films. Results indicate that this dose-mapping technique is a powerful method of evaluating a variety of radiation fields of commercial radiation sources, with promising applications as a means of dose validation and quality control. (Author)

  14. 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.

  15. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

    In order to verify the materials as an important artificial barrier for canister of vitrified high-level waste from spent fuel reprocessing, data and reports were researched on corrosion resistance of the materials under conditions from glass form production to final disposal. Then, in this report, investigated subjects, improvement methods and future subjects are reviewed. It has become clear that there would be no problem on the inside and outside corrosion of the canister during glass production, but long term corrosion and radiation effect tests and the vitrification methods would be subjects in future on interim storage and final disposal conditions. (author)

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

    International Nuclear Information System (INIS)

    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

  17. Chemical stability of copper-canisters in deep repository

    International Nuclear Information System (INIS)

    The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper-iron canisters and placed deep into the bedrock. The copper wall of the canister provides a long-time shield against corrosion, preventing the high-level nuclear fuel from contact with ground water. In the report, stability of metallic copper and its possible corrosion reactions in the conditions of deep bedrock are evaluated by means of thermo-dynamic calculations. (90 refs., 28 figs., 11 tabs.)

  18. 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.

  19. 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.

  20. Interaction between rock, bentonite buffer and canister. FEM calculations of some mechanical effects on the canister in different disposal concepts

    International Nuclear Information System (INIS)

    An important task of the buffer of highly compacted bentonite is to offer a mechanical protection to the canister. This role has been investigated by a number of finite element calculations using the complex elasto plastic material models for the bentonite that have been developed on the basis of laboratory tests and adapted to the code ABAQUS. The following main functions and scenarios have been investigated for some different canister types and repository concepts: - The effect of the water and swelling pressure, - The effect of a rock shear perpendicular to the canister axis, - The effect of creep in the copper after a rock shear displacement, - The thermomechanical effects when an initially saturated buffer is used

  1. Design basis for the copper/steel canister. Stage five. Final report

    International Nuclear Information System (INIS)

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste in the Swedish Program, has been studied by the present author from the points of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress that has been made between May-1-1998 and April-30-1999 and the result of further literature studies. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. The nodular iron that was selected has been the subject of casting trials at several foundries. Early trials, using uphill feeding, met with limited success owing to difficulties feeding during solidification. Lessons from this trial led to a modification to the casting design to include extra cores that have the effect of reducing the need for feeding in the heaviest sections. Results using the new design and direct (downhill) casting are very promising. Castings appear to be sound and mechanical test results cast-on bars are within specification. Tensile test results from specimens cut from the casting have reduced ductility compared with the cast-on bars and this may be evidence of microstructural variations within the casting. The material specified for the overpack is OF (Oxygen Free) copper with 50 ppm of phosphorus added. Concentration limits have now been placed on impurity elements which are below those allowed in the OF specification. All current trials are using material from Outokompu produced from cathode on their OF(E) line, which delivers total impurity levels of less than 30 ppm excluding silver and phosphorus. The phosphorus addition is made using a master alloy added to the launder and this does not give good control of phosphorus level either within or between castings. Phosphorus is added to improve creep rates and creep strain to failure. The level is limited to 50 ppm in order to avoid difficulties, which it might

  2. Study on long-term corrosion behavior of high corrosion resistant metal overpack under reducing condition

    International Nuclear Information System (INIS)

    For repository container material of high-level radioactive wastes, titanium, nickel-based alloys, etc. have been investigated as high corrosion resistant metal. Titanium has excellent corrosion resistance under high chloride and oxidizing conditions and has many applications in general industries. However it has a possibility to absorb hydrogen generated by water reduction and cause hydrogen embrittlement under reducing condition of the repository. In this study, experimental investigation was carried out on hydrogen absorption behavior of titanium and influencing factors under reducing condition. In addition, previous studies were searched on the corrosion resistant material other than titanium. (1) Electrochemical acceleration tests of titanium were carried out to apply cathodic electric charge equivalent to the corrosion for 1000 years under reducing condition. The effects of parameters, processing rate, heat treatment conditions, applied stress and solution pH, on the hydrogen absorption rate were evaluated. (2) Sealed ampoule type immersion tests were conducted under reducing condition. Effects of pH on the hydrogen absorption rate were evaluated and furthermore, the surface hydrogen concentration was analyzed. (3) Hydrogen absorption/embrittlement models of titanium overpack were discussed and hydrogen concentration distribution after 1000 years was predicted. (4) Previous studies on corrosion behavior of high corrosion resistant alloys other than titanium were searched and organized. In addition, an investigation was conducted on the selection of waste package materials in the United States. (author)

  3. Resin Liner Recovery and Over-Packing at Ontario Power Generation's Western Waste Management Facility

    International Nuclear Information System (INIS)

    Spent resins generated from Ontario Power Generation (OPG)'s and Bruce Power's Candu reactor operations are stored at OPG's Western Waste Management Facility in Kincardine, Ontario, Canada. The older resins are contained in 3 m3 epoxy-coated cylindrical carbon steel containers known as resin liners. The liners are stored in a stacked configuration within cylindrical in-ground containers. Previous studies indicated evidence of unacceptable liner wall corrosion and the potential for eventual leakage of resin from the liners. Based on this, OPG elected to re-package the majority of the resin liners into stainless steel over-packs. A contract for this work was awarded to a project team consisting of Duratek of Canada, Kinectrics, Inc. and E.S. Fox. This paper provides an overall summary of project activities focusing on the effectiveness of the equipment utilized and the soundness of the developed programs, plans and procedures. Specific information is provided on key aspects of the project and the overall achievement of project goals. (authors)

  4. Study on the methods for analysis of the chemical poison in canister by neutron activity

    International Nuclear Information System (INIS)

    The method that is used to analyse the poison gases in canister by neutron activity is proposed. Through theory analysis and experimental measurement, the feasibility for analysis of the poison gases in a canister by neutron activity has been demonstrated, and it is proved that the method itself do not result in radioactive problem to use again the canister. (authors)

  5. Corrosion resistance of a copper canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    The report presents an evaluation of copper as canister material for spent nuclear fuel. The evaluation is made from the viewpoint of corrosion and applies to a concept of 1977. Supplementary corrosion studies have been performed. The report includes 9 appendices which deal with experimental data. (G.B.)

  6. Analysis of water from K west basin canisters (second campaign)

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

    Gas and liquid samples have been obtained 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. The data will provide 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 in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results for the gas and liquid samples of the first campaign have been reported (Trimble 1995a; Trimble 1995b; Trimble 1996a; Trimble 1996b). An analysis of cesium-137 (137CS ) data from the second campaign samples was reported (Trimble and Welsh 1997), and the gas sample results are documented in Trimble 1997. This report documents the results of all analytes of liquid samples from the second campaign.

  7. 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.

  8. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    Several concepts are described for the interim (20 to 30 years) storage of canisters containing high level waste, cladding waste, and intermediate level-TRU wastes. It includes requirements, ground rules and assumptions for the entire storage pilot plant. Concepts are generally evaluated and the most promising are selected for additional work. Follow-on recommendations are made

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

    International Nuclear Information System (INIS)

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains ampersand 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

  10. Effects of glacial meltwater on corrosion of copper canisters

    International Nuclear Information System (INIS)

    The composition of glacial meltwater and its reactions in the bedrock are examined. The evidences that there are or should be from past intrusions of glacial meltwater and oxygen deep in the bedrock are also considered. The study is concluded with an evaluation of the potential effects of oxygenated meltwater on the corrosion of copper canisters. (46 refs., 3 figs., 2 tabs.)

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

    International Nuclear Information System (INIS)

    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

  12. Chemical durability of copper canisters under crystalline bedrock repository conditions

    International Nuclear Information System (INIS)

    In the Swedish waste management program, the copper canister is expected to provide containment of the radionuclides for a very long time, perhaps millions of years. The purpose of the present paper, is to analyze prerequisites for assessments of corrosion lifetimes for copper canisters. The analysis is based on compilations of literature from the following areas: chemical literature on copper and copper corrosion, mineralogical literature with emphasis on the stability of copper in near surface environments, and chemical and mineralogical literature with emphasis on the stabilities and thermodynamics of species and phases that may exist in a repository environment. Three main types of situations are identified: (1) under oxidizing and low chloride conditions, passivating oxide type of layers may form on the copper surface; (2) under oxidizing and high chloride conditions, the species formed may all be dissolved; and (3) under reducing conditions, non-passivating sulfide type layer may form on the copper surface. Considerable variability and uncertainty exists regarding the chemical environment for the canister, especially in certain scenarios. Thus, the mechanisms for corrosion can be expected to differ greatly for different situations. The lifetime of a thick-walled copper canister subjected to general corrosion appears to be long for most reasonable chemistries. (It is assumed that the canister has no defects from manufacturing and that the bentonite buffer is intact). Localized corrosion may appear for types (1) and (3) above but the mechanisms are widely different in character. The penetration caused by localized corrosion can be expected to be very sensitive to details in the chemistry

  13. 42 CFR 84.1153 - Dust, fume, mist, and smoke tests; canister bench tests; gas masks canisters containing filters...

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Dust, fume, mist, and smoke tests; canister bench... RESEARCH AND RELATED ACTIVITIES APPROVAL OF RESPIRATORY PROTECTIVE DEVICES Dust, Fume, and Mist; Pesticide; Paint Spray; Powered Air-Purifying High Efficiency Respirators and Combination Gas Masks § 84.1153...

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

    International Nuclear Information System (INIS)

    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

  15. Heat propagation from a radioactive waste repository. SKB 91 reference canister

    International Nuclear Information System (INIS)

    A study of heat propagation around a hypothetical radioactive waste repository is presented. The investigated flow domain was limited to a quarter of the flow domain around a single canister due to symmetry by vertical planes passing through the centre of the canister, half distance between the adjacent tunnels and the adjacent canisters. Strictly speaking, such an approach is applicable to a repository of infinite extent. However, from a practical point of view this assumption applies to all canisters but the ones close to the edge of the repository. The following different material regions were considered: (a.) Canister containing the spent fuel, (b.) Buffer (bentonite) around the canister, (c.) Backfilled (mixture of bentonite and sand) tunnels, and (d.) host Rock. The canister material was presented by a 'homogenized' medium obtained by weighted averaging of the main constituents of the canister, viz. spent fuel, copper and lead. A geothermal gradient of 13 degrees C/km was assumed. The initial heat effect per canister was 1066 W. The total vertical extent of the flow domain considered was about 1500 meters. The base case, with 6.2 m canister spacing and 30 m tunnel spacing, resulted in a maximum temperature at the canister/buffer interface of about 66 degrees C (corresponding to a temperature rise of about 54 degrees C), and about 50 degrees C (about 38 degrees C temperature rise) in the rock. (au)

  16. Pressure tests of two KBS-3 canister mock-ups

    International Nuclear Information System (INIS)

    The Swedish concept for geological disposal of spent nuclear fuel, the so-called KBS-3 concept, relies on a multibarrier system with the copper/cast iron canister as the first barrier. The canister is designed to retain its integrity for at least 100,000 years, which means that future glaciations need to be considered. A 3 km thick ice block together with hydrostatic pressure from groundwater and swelling of the buffer material would produce hydrostatic compressive stresses of maximum 44 MPa (440 bar). Although the canister is loaded globally in compression, tensile stresses develop at fuel channel surface with increasing load. Tensile tests of the insert material in the development phase of the KBS-3 canister indicated a large scatter and relatively low values of the inserts' ductility. An important issue was whether this could lead to mechanical failure of canisters at the 44 MPa iso-static load either by plastic collapse or fracture from the defects in the regions with tensile stresses. SKB therefore initiated a project together with the European commission's Joint Research Centre (JRC) Institute of Energy in Petten and a number of Swedish partners to evaluate the probability of mechanical failure during glaciation. Three inserts manufactured by different Swedish foundries and referred to as 1, 125 and 126 were used in the project. A large statistical test programme was developed to determine statistical distributions of various material parameters and defect distributions. These data were subsequently used in probabilistic analysis to determine the probability for local plastic collapse or fracture. The main conclusion was that the failure probability is extremely low at the design load (44 MPa) provided some basic geometrical requirements are fulfilled. In parallel to the statistical test programme and the associated analysis, the group decided also to perform two pressure tests of canister mock-ups to demonstrate the actual safety margins. The fractographic

  17. Drying tests conducted on Three Mile Island fuel canisters containing simulated debris

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A.J.

    1995-12-31

    Drying tests were conducted on TMI-2 fuel canisters filled with simulated core debris. During these tests, canisters were dried by heating externally by a heating blanket while simultaneously purging the canisters` interior with hot, dry nitrogen. Canister drying was found to be dominated by moisture retention properties of a concrete filler material (LICON) used for geometry control. This material extends the drying process 10 days or more beyond what would be required were it not there. The LICON resides in a nonpurgeable chamber separate from the core debris, and because of this configuration, dew point measurements on the exhaust stream do not provide a good indication of the dew point in the canisters. If the canisters are not dried, but rather just dewatered, 140-240 lb of water (not including the LICON water of hydration) will remain in each canister, approximately 50-110 lb of which is pore water in the LICON and the remainder unbound water.

  18. Qualification of final closure for disposal container I - applicability of TIG and EBW for overpack welding

    International Nuclear Information System (INIS)

    Regarding the final sealing technique of the overpack using carbon steel, one of the candidate materials for the disposal container in the geological disposal of high-level radioactive waste in Japan, welding tests were conducted using TIG (GTAW), a typical arc welding process, and electron beam welding (EBW), a high-energy beam welding process. The purpose of the tests was to evaluate the applicability, the scope of the applications and the conditions for the application of the existing techniques; while also examining the welding conditions and the weld quality. Regarding TIG, the optimum welding conditions (the conditions pertaining to the welding procedures and the groove geometry) were checked by using a specimen with a plate thickness of 50 mm, and then circumferential welding tests were conducted for cylindrical specimens with a groove depth of 100 mm and 150 mm. Radiographic testing showed that there was no significant weld defect in the weld and that the welding characteristics were satisfactory. The results of the test of the mechanical properties of the joint were also satisfactory. Measurement of the temperature distribution and the residual stress distribution at the time of the welding was conducted for an evaluation of the residual stress caused by the welding, and an appropriate residual stress analysis method was developed, which confirmed the generation of tensile stress along the circumferential direction of the weld. Then it was pointed out that a necessity of further consideration of how to reduce the stress and to examine the influence that residual stress has on corrosion property. The goal in the EBW test was to achieve a one-pass full penetration welding process for 190 mm while conducting a partial penetration welding test for a welding depth of 80 mm. Subsequent radiographic testing confirmed that there was no significant weld defect. (orig.)

  19. Status of fabrication of square-format masks for extreme-ultraviolet lithography (EUVL) at the MCoC

    Science.gov (United States)

    Racette, Kenneth C.; Williams, Carey T.; Fisch, Emily; Kindt, Louis; Lawliss, Mark; Ackel, Robin; Lercel, Michael J.

    2002-07-01

    Fabricating masks for extreme ultraviolet lithography is challenging. The high absorption of most materials at 13.4 nm and the small critical dimension (45 nm) at the target insertion node force many new features, including reflective mask design, new film choices, and stringent defect specifications. Fabrication of these masks requires the formation and patterning of both a repair buffer layer and an EUV absorber layer on top of a molybdenum/silicon multi-layer substrate. IBM and Photronics have been engaged in developing mask processing technology for x-ray, electron beam projection and extreme ultraviolet lithographies at the Next Generation Lithography Mask Center of Competency (NGL-MCoC) within IBM's mask facility at Essex Junction, Vermont. This paper describes recent results of mask fabrication on 6 x 6 x 1/4 inch EUVL substrates (quartz with molybdenum silicon multi-layers) at the MCoC. Masks fabricated with high and low-stress chromium and externally deposited chromium absorber films are compared. In particular, etch characteristics, image size, image placement, line edge roughness, and defect levels are presented and compared. Understanding the influence of the absorber film characteristics on these parameters will enable us to optimize the effectiveness of a given absorber film or to select acceptable alternatives.

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

    International Nuclear Information System (INIS)

    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.)

  1. Safety evaluation for the inner canister closure station

    International Nuclear Information System (INIS)

    The Inner Canister Closure Station (ICCS), built by Remote Technology Corporation, will be operability tested. The ICCS is used to remotely leak test Inner Canister Closures (ICC's) and replace ICC's that are not water tight. After operability testing, the ICCS will be inspected and sent to the 717-F mock-up shop for remotability demonstration and dimensional checks, then installed in the Vitrification Building, 221-S. An analysis of potential safety hazards, equipment safety features, and procedural controls indicates that the ICCS can be operated without undue hazard to employees or to the public. A safety inspection and a new equipment inspection will be held before operation to verify that the ICCS meets Savannah River Site safety requirements. 4 refs., 6 figs

  2. Multi-purpose canister system evaluation: A systems engineering approach

    International Nuclear Information System (INIS)

    This report summarizes Department of Energy (DOE) efforts to investigate various container systems for handling, transporting, storing, and disposing of spent nuclear fuel (SNF) assemblies in the Civilian Radioactive Waste Management System (CRWMS). The primary goal of DOE's investigations was to select a container technology that could handle the vast majority of commercial SNF at a reasonable cost, while ensuring the safety of the public and protecting the environment. Several alternative cask and canister concepts were evaluated for SNF assembly packaging to determine the most suitable concept. Of these alternatives, the multi-purpose canister (MPC) system was determined to be the most suitable. Based on the results of these evaluations, the decision was made to proceed with design and certification of the MPC system. A decision to fabricate and deploy MPCs will be made after further studies and preparation of an environmental impact statement

  3. 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.

  4. BRIC-60: Biological Research in Canisters (BRIC)-60

    Science.gov (United States)

    Richards, Stephanie E. (Compiler); Levine, Howard G.; Romero, Vergel

    2016-01-01

    The Biological Research in Canisters (BRIC) is an anodized-aluminum cylinder used to provide passive stowage for investigations evaluating the effects of space flight on small organisms. Specimens flown in the BRIC 60 mm petri dish (BRIC-60) hardware include Lycoperscion esculentum (tomato), Arabidopsis thaliana (thale cress), Glycine max (soybean) seedlings, Physarum polycephalum (slime mold) cells, Pothetria dispar (gypsy moth) eggs and Ceratodon purpureus (moss).

  5. Analysis of probability of defects in the disposal canisters

    International Nuclear Information System (INIS)

    This report presents a probability model for the reliability of the spent nuclear waste final disposal canister. Reliability means here that the welding of the canister lid has no critical defects from the long-term safety point of view. From the reliability point of view, both the reliability of the welding process (that no critical defects will be born) and the non-destructive testing (NDT) process (all critical defects will be detected) are equally important. In the probability model, critical defects in a weld were simplified into a few types. Also the possibility of human errors in the NDT process was taken into account in a simple manner. At this moment there is very little representative data to determine the reliability of welding and also the data on NDT is not well suited for the needs of this study. Therefore calculations presented here are based on expert judgements and on several assumptions that have not been verified yet. The Bayesian probability model shows the importance of the uncertainty in the estimation of the reliability parameters. The effect of uncertainty is that the probability distribution of the number of defective canisters becomes flat for larger numbers of canisters compared to the binomial probability distribution in case of known parameter values. In order to reduce the uncertainty, more information is needed from both the reliability of the welding and NDT processes. It would also be important to analyse the role of human factors in these processes since their role is not reflected in typical test data which is used to estimate 'normal process variation'.The reported model should be seen as a tool to quantify the roles of different methods and procedures in the weld inspection process. (orig.)

  6. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    International Nuclear Information System (INIS)

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  7. CLASSIFICATION OF THE MGR CANISTERED SNF DISPOSAL CONTAINER SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) canistered spent nuclear fuel disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333PY ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  8. Development of ultrasonic immersion inspection technique for spent fuel canisters

    International Nuclear Information System (INIS)

    This report summarizes ultrasonic nondestructive testing development for metal matrix supported spent fuel disposal canisters. The work has concentated in two areas: inspection for lack of bond at the shell/matrix interface and inspection for voids in the matrix. The capabilities and limitations of these techniques have been fully established. Unbonded areas as small as 4 mm in diameter and voids 6 mm in diameter, 25 mm deep in the matrix, can readily be detected

  9. Canister displacement in KBS-3V. A theoretical study

    International Nuclear Information System (INIS)

    The vertical displacement of the canister in the KBS-3V concept has been studied in a number of consolidation and creep calculations using the FE-program ABAQUS. The creep model used for the calculations is based on Singh-Mitchell's creep theory, which has been adapted to and verified for the buffer material MX-80 in earlier tests. A porous elastic model with Drucker-Prager plasticity has been used for the consolidation calculations. For simplicity the buffer has been assumed to be water saturated from start. In one set of calculations only the consolidation and creep in the buffer without considering the interaction with the backfill was studied. In the other set of calculations the interaction with the backfill was included for a backfill consisting of an in situ compacted mixture of 30% bentonite and 70% crushed rock. The motivation to also study the behaviour of the buffer alone was that the final choice of backfill material and backfilling technique is not made yet so that set of calculations simulates a backfill that has identical properties with the buffer. The two cases represent two extreme cases, one with a backfill that has a low stiffness and the lowest allowable swelling pressure and one that has the highest possible swelling pressure and stiffness. The base cases in the calculations correspond to the final average density at saturation of 2,000 kg/m3 with the expected swelling pressure of 7 MPa in a buffer. In order to study the sensitivity of the system to loss in bentonite mass and swelling pressure seven additional calculations were done with reduced swelling pressure down to 80 kPa corresponding to a density at water saturation of about 1,500 kg/m3. The calculations included two stages, where the first stage models the swelling and consolidation that takes place in order for the buffer to reach force equilibrium. This stage takes place during the saturation phase and the subsequent consolidation/swelling phase. The second stage models the

  10. 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.

  11. Test report for the Sample Transfer Canister system

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, B.D.

    1998-03-04

    The Sample Transfer Canister will be used by the Waste Receiving and Processing Facility (WRAP) for the transport of small quantity liquid samples that meet the definition of a limited quantity radioactive material, and may also be corrosive and/or flammable. Transport of the system will typically be north of the Wye Barricade between WRAP and the 222-S Laboratory. The samples are intended to conform to the US Department of Transportation (DOT) regulation 49 CFR 1 73.4, ``Exceptions for small quantities.`` The regulations require prototype testing of the package to demonstrate the effectiveness of the packaging system. The test procedure consisted of one 24-hour compression test and five drop tests of various orientations onto an unyielding drop pad. The testing of the Sample Transfer Canister System was performed between February 16, 1998 and February 25, 1998. The results of the testing concluded that the Sample Transfer Canister System successfully met the testing requirements with certain modifications to the original system. The modifications included replacing the original eight flange screws which were cold rolled 316 stainless steel with greater strength grade 8 high carbon-carbon steel screws, replacing the initial two glass receptacles with a better performing single glass receptacle which proved not to leak during testing, and adding more bubble wrap as extra padding.

  12. Test programs conducted in support of high-level waste canister fabrication using radioactively contaminated steel

    International Nuclear Information System (INIS)

    The Canister Fabrication Development Activity (CFDA) was developed at the INEL to investigate the potential of fabricating high-level waste (HLW) canisters from radioactively contaminated stainless steel. Metal melting and forming processes were evaluated, and centrifugal casting was the method ultimately chosen for the process to fabricate the cylindrical portion of the HLW canister. Test programs were conducted to determine if a centrifugally cast (CF-3) stainless steel canister is equivalent to a wrought 304L stainless steel canister and to determine what problems might result from melting, casting, machining, and utilizing canisters fabricated from radioactively contaminated steel. A survey was also made of the radioactively contaminated stainless steel volumes in the United States to determine a source of steel for fabrication of the canisters. The results of the survey showed that there are up to 30,000 tons of radioactively contaminated stainless steel that could be available over the next 25 years. The results of these tests showed that centrifugally cast canisters are an acceptable alternative to wrought canisters and that HLW canisters can be successfully fabricated from radioactively contaminated steel

  13. Thermal Dimensioning of SiC Canister Applied A-KRS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Inyoung; Choi, Heuijoo; Yoo, Malgobalgebitnala [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Reducing toxicity and volume of SNF and reusing valuable fissile materials, pyro-processing connected with SFR is under-developing. The A-KRS is composed of 1 cm thick copper cold-spray-coated cast iron canisters, buffer blocks, disposal holes and disposal tunnels, etc. To manufacture disposal canisters, massive un-reusable copper and iron resources are required. Recently, SiC which has high thermal conductivity and good mechanical properties is investigated as a substitute material of metal canister to save metal resources. In this study, thermal performance of SiC canister is investigated and thermal dimensioning of SiC canister applied A-KRS is conducted to estimate thermal applicability of SiC canister in repository. In this study, thermal applicability of SiC as a substitute material of copper and cast iron canister is assessed. Due to higher thermal conductivity of SiC, calculated maximum temperature of SiC applied system is lower than original metal canister applied system and estimated minimum disposal hole pitch of SiC canister system is narrower than metal canister system. But decrease of distance between disposal hole pitch by adopting SiC canister is negligible considering engineering and safety margin. As a result, it is confirmed that SiC could be used as a substitute materials of metal in respect of thermal aspect. To apply SiC canister in deep geological repository, however, thermal-mechanical assessment need to be conducted as future studies. Especially thermally induced stress and intactness of canister must be estimated because SiC is fragile material and its thermal conductivity is highly dependent on temperature.

  14. Thermal Dimensioning of SiC Canister Applied A-KRS

    International Nuclear Information System (INIS)

    Reducing toxicity and volume of SNF and reusing valuable fissile materials, pyro-processing connected with SFR is under-developing. The A-KRS is composed of 1 cm thick copper cold-spray-coated cast iron canisters, buffer blocks, disposal holes and disposal tunnels, etc. To manufacture disposal canisters, massive un-reusable copper and iron resources are required. Recently, SiC which has high thermal conductivity and good mechanical properties is investigated as a substitute material of metal canister to save metal resources. In this study, thermal performance of SiC canister is investigated and thermal dimensioning of SiC canister applied A-KRS is conducted to estimate thermal applicability of SiC canister in repository. In this study, thermal applicability of SiC as a substitute material of copper and cast iron canister is assessed. Due to higher thermal conductivity of SiC, calculated maximum temperature of SiC applied system is lower than original metal canister applied system and estimated minimum disposal hole pitch of SiC canister system is narrower than metal canister system. But decrease of distance between disposal hole pitch by adopting SiC canister is negligible considering engineering and safety margin. As a result, it is confirmed that SiC could be used as a substitute materials of metal in respect of thermal aspect. To apply SiC canister in deep geological repository, however, thermal-mechanical assessment need to be conducted as future studies. Especially thermally induced stress and intactness of canister must be estimated because SiC is fragile material and its thermal conductivity is highly dependent on temperature

  15. Stress analysis of high-level waste canisters: methods, applications, and design data

    International Nuclear Information System (INIS)

    An overview of stress analysis methods, structural design procedures, and design data is presented for canisters used to package solidified wastes, particularly borosilicate glass. In addition, waste processing, canister materials, fabrication and inspection methods, and performance testing are summarized. Sources of stress in canisters are lifting and handling loads, internal pressure, high-temperature filling operations, transient heating and cooling, differential thermal expansions of canisters and glass, and impact loadings from low-probability accidents. Results of case studies that illustrate applicable methods of stress analyses are presented for these sources of stress. Existing sections of ASME Boiler and Pressure Vessel Code are applicable to canister fabrication, but the code does not cover many aspects of canister service loadings. Specialized criteria for minimum wall thicknesses to sustain filling stresses are proposed in this report. Results of a test program to measure the creep strength of candidate canister materials are described. Methods to predict residual stresses in the walls of waste canisters are described; predicted residual stress levels agree with measured stress levels. The consequences of these residual stresses are reviewed, and stress-corrosion cracking is identified as the mode of canister failure affected by residual stresses. Canister-closure design is covered in detail, particularly the welding and inspection of the final closure seal-weld. It is shown that the methods of fracture mechanics and fatigue-crack-growth analyses are valuable tools for evaluating the performance of closure welds in the presence of crack-like defects. Canister performance in process trials at PNL shows the ability of canisters to survive high temperatures and loadings during processing. Impact tests show that a suitably designed canister can sustain severe impacts without loss of intergrity

  16. Syntheses, structure and magnetic properties of two vanadate garnets Ca5M4V6O24 (M=Co, Ni)

    International Nuclear Information System (INIS)

    Two vanadate compounds Ca5M4V6O24 (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 Ca5M4V6O24 (M=Co, Ni) exhibit similar magnetic behaviors, in which Ca5Co4V6O24 possesses an antiferromagnetic ordering at TN=~6 K while Ca5Ni4V6O24 shows an antiferromagnetic ordering at TN=~7 K. - Graphical abstract: Garnet vanadate compounds Ca5M4V6O24 (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 Ca5M4V6O24 (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

  17. Test manufacturing of copper canisters with cast inserts. Assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, C.G

    1998-08-01

    The current design of canisters for the deep repository for spent nuclear fuel consists of an outer corrosion-protective copper casing in the form of a tubular section with lid and bottom and an inner pressure-resistant insert. The insert is designed to be manufactured by casting and inside are channels in which the fuel assemblies are to be placed. Over the last years, a number of full-scale manufacturing tests of all canister components have been carried out. The purpose has been to determine and develop the best manufacturing technique and to establish long-term contacts with the best suppliers of material and technology. Part of the work has involved the developing and implementing of a quality assurance system in accordance with ISO 9001, covering the whole chain from suppliers of material up to and including the delivery of assembled canisters. This report consists of a description of the design of the canister together with current drawings and complementary technical specifications stipulating, among other things, requirements placed on different materials. The different manufacturing methods that have been used are also described and commented on in both text and illustrations. For the manufacturing of copper tubes, the roll-forming of rolled plate to tube halves and longitudinal welding is a method that has been tested on a relatively large number of tubes by now, and that probably can be developed into a functioning production method. However, the very promising outcome of performed tests on seamless tube manufacturing, has resulted in a change in direction in tube manufacturing, focusing on continued testing of extrusion as well as pierce and draw processing in the immediate future. In connection with ongoing operations, new manufacturing tests of tubes with less material thickness will be carried out. Test manufacturing of cast inserts has resulted in the choice of nodular iron as material in the continued work. This improvement in design has resulted

  18. Test manufacturing of copper canisters with cast inserts. Assessment report

    International Nuclear Information System (INIS)

    The current design of canisters for the deep repository for spent nuclear fuel consists of an outer corrosion-protective copper casing in the form of a tubular section with lid and bottom and an inner pressure-resistant insert. The insert is designed to be manufactured by casting and inside are channels in which the fuel assemblies are to be placed. Over the last years, a number of full-scale manufacturing tests of all canister components have been carried out. The purpose has been to determine and develop the best manufacturing technique and to establish long-term contacts with the best suppliers of material and technology. Part of the work has involved the developing and implementing of a quality assurance system in accordance with ISO 9001, covering the whole chain from suppliers of material up to and including the delivery of assembled canisters. This report consists of a description of the design of the canister together with current drawings and complementary technical specifications stipulating, among other things, requirements placed on different materials. The different manufacturing methods that have been used are also described and commented on in both text and illustrations. For the manufacturing of copper tubes, the roll-forming of rolled plate to tube halves and longitudinal welding is a method that has been tested on a relatively large number of tubes by now, and that probably can be developed into a functioning production method. However, the very promising outcome of performed tests on seamless tube manufacturing, has resulted in a change in direction in tube manufacturing, focusing on continued testing of extrusion as well as pierce and draw processing in the immediate future. In connection with ongoing operations, new manufacturing tests of tubes with less material thickness will be carried out. Test manufacturing of cast inserts has resulted in the choice of nodular iron as material in the continued work. This improvement in design has resulted

  19. Sorption behavior of some TRUs and FPs onto corrosion products of overpacks for geological disposal of high-level radwaste

    International Nuclear Information System (INIS)

    The sorption of TRUs and FPs on magnetite and goethite was studied to basically examine the contribution of the iron minerals to the retardation of radionuclide migration. The sorption behavior was clarified using a batch technique under various redox conditions and solid/liquid ratios with the ionic strength of 0.1 M (NaClO4) at the temperature 25degC. TRU (Np, Pu and Am) showed large distribution coefficients ranging between 5 and 200 m3/kg. The sorption of Np, Pu and Eu on the mineral was enhanced under the presence of carbonates. The preliminary calculation using Kd model showed that nuclide concentration could be decreased by more than two orders when 0.001 of the corroded overpack is effective to sorption. RAPRAN calculation was performed considering the sorption effect of corroded iron assuming time dependent overpack corrosion. The release of the TRUs from the edge of the engineered barrier system was reduced due to the sorption onto the corroded iron. (author)

  20. Kinetic modelling of bentonite-canister interaction. Long-term predictions of copper canister corrosion under oxic and anoxic conditions

    International Nuclear Information System (INIS)

    A new modelling approach for canister corrosion which emphasises chemical processes and diffusion at the bentonite-canister interface is presented. From the geochemical boundary conditions corrosion rates for both an anoxic case and an oxic case are derived and uncertainties thereof are estimated via sensitivity analyses. Time scales of corrosion are assessed by including calculations of the evolution of redox potential in the near field and pitting corrosion. This indicates realistic corrosion depths in the range of 10-7 and 4*10-5 mm/yr, respectively for anoxic and oxic corrosion. Taking conservative estimates, depths are increased by a factor of about 200 for both cases. From these predictions it is suggested that copper canister corrosion does not constitute a problem for repository safety, although certain factors such as temperature and radiolysis have not been explicitly included. The possible effect of bacterial processes on corrosion should be further investigated as it might enhance locally the described redox process. 35 refs, 11 figs, 6 tabs

  1. Thermo-hydro-mechanical mode of canister retrieval test

    International Nuclear Information System (INIS)

    Document available in extended abstract form only. The Canister Retrieval Tests (CRT) is a full scale in situ experiment performed by SKB at Aespoe Laboratory. The experiment involves placing a canister equipped with electrical heaters inside of a deposition hole bored in Aespoe diorite. The deposition hole is 8.55 metres deep and has a diameter of 1.76 metres. The space between canister and the hole is filled with a MX-80 bentonite buffer. The bentonite buffer was installed in form of blocks and rings of bentonite. At the top of the canister bentonite bricks occupy the volume between the canister top surface and the bottom surface of the plug. Due to the bentonite ring size there are two gaps; once between canister and buffer which was left empty and another one between buffer and rock that was filled with bentonite pellets. The top of the hole was sealed with a retaining plug composed of concrete and a steel plate. The plug was secured against heave caused by the swelling clay with nine cables anchored in the rock. An artificial pressurised saturation system was used because the supply of water from the rock was judged to be insufficient for saturating the buffer in a feasible time. A large number of instruments were installed to monitor the test as follows: - Canister - temperature and strain. - Rock mass - temperature and stress. - Retaining system - force and displacement. - Buffer - temperature, relative humidity, pore pressure and total pressure. After dismantling the tests the final dry density and water content of bentonite and pellets were measured. The comprehensive record of the Thermo-Hydro-Mechanical (THM) processes in the buffer give the possibility to investigate theoretical formulations and models, since the results of THM analyses can be checked against experimental data. As part of the European project THERESA, a 2-D axisymmetric model simulation of CRT bas been carried out. Some of the main objectives of this simulation are the study of the

  2. On the origin of the conductance asymmetry in CeMIn5 (M=Co, Rh, Ir)

    International Nuclear Information System (INIS)

    Asymmetric differential conductance has been frequently observed in heavy fermion point-contact junctions. We report such data obtained from the Ce-based 1-1-5 compounds CeMIn5 (M=Co, Rh, Ir). Apart from characteristics due to superconductivity or antiferromagnetism, a striking common feature is an asymmetry in the background conductance, which shows nontrivial temperature and voltage dependencies. These behaviors cannot be explained by the local heating model combined with large Seebeck effect in heavy fermions. We propose that a Fano-like interference may cause the asymmetry. The interference can occur between two conductance channels, one into the conduction band and the other into the heavy electron band formed by the hybridization of conduction electrons with localized f-electrons.

  3. Perovskite LaPbMSbO6 (M=Co, Ni): Structural distortion, magnetic and dielectric properties

    International Nuclear Information System (INIS)

    The B-site ordered double perovskite oxides LaPbMSbO6 (M=Co, Ni) have been synthesized via the modified Sol–Gel precursor two-step route. Rietveld refinements reveal strong abnormal structural distortion and BO6 octahedral deformation appearing along the ab plane. Owing to the cooperative Jahn–Teller effect of Co2+ and Pb2+ ions, the Co-related compound exhibits almost complete Co2+–Sb5+ order. For magnetic properties, spin-canted antiferromagnetic state with high extent of magnetic frustration is confirmed. The Ni-related compound presents heavier magnetic frustration for introducing tiny disorder on site occupation accompanied with valence state and further enhancing the complexity of magnetic competition. Dielectric measurements present a considerable temperature dependent dielectric relaxation with great dc-like loss feature in the LaPbCoSbO6. For LaPbNiSbO6, however, the permittivity with low dielectric loss is shown to be insensitive to either temperature or frequency. The corresponding electronic active energy manifests that the weakly bounded 3d-electron is prone to hop in a more distorted Co–Sb sublattice. - Graphical abstract: XRD Rietveld refinement result of LaPbCoSbO6 presented a large BO6 octahedral distortion along the ab plane. Based upon the variations from Co–O–Sb bond angles, a fierce competition from many extended magnetic coupling routes (M–O–O–M) would induce a considerably large magnetic frustration and electron hopping restriction. - Highlights: • Highly ordered LaPbMSbO6 (M=Co, Ni) were synthesized. • Abnormal structural distortion appeared in the ab plane. • Strong magnetic frustration was confirmed via M2+–O–O–M2+ route. • Dielectric measurements presented a large difference between Co and Ni samples. • 3d-electronic structure determines lattice distortion and physical properties

  4. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which results in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical. (au)

  5. 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.

  6. SITE-94. CAMEO: A model of mass-transport limited general corrosion of copper canisters

    International Nuclear Information System (INIS)

    This report describes the technical basis for the CAMEO code, which models the general, uniform corrosion of a copper canister either by transport of corrodants to the canister, or by transport of corrosion products away from the canister. According to the current Swedish concept for final disposal of spent nuclear fuels, extremely long containment times are achieved by thick (60-100 mm) copper canisters. Each canister is surrounded by a compacted bentonite buffer, located in a saturated, crystalline rock at a depth of around 500 m below ground level. Three diffusive transport-limited cases are identified for general, uniform corrosion of copper: General corrosion rate-limited by diffusive mass-transport of sulphide to the canister surface under reducing conditions; General corrosion rate-limited by diffusive mass-transport of oxygen to the canister surface under mildly oxidizing conditions; General corrosion rate-limited by diffusive mass-transport of copper chloride away from the canister surface under highly oxidizing conditions. The CAMEO code includes general corrosion models for each of the above three processes. CAMEO is based on the well-tested CALIBRE code previously developed as a finite-difference, mass-transfer analysis code for the SKI to evaluate long-term radionuclide release and transport in the near-field. A series of scoping calculations for the general, uniform corrosion of a reference copper canister are presented

  7. COMSOL Multiphysics Model For DWPF Canister Filling, Revision 1

    International Nuclear Information System (INIS)

    This revision is an extension of the COMSOL Multiphysics model previously developed and documented to simulate the temperatures of the glass during pouring a Defense Waste Processing Facility (DWPF) canister. In that report the COMSOL Multiphysics model used a lumped heat loss term derived from experimental thermocouple data based on a nominal pour rate of 228 lbs./hr. As such, the model developed using the lumped heat loss term had limited application without additional experimental data. Therefore, the COMSOL Multiphysics model was modified to simulate glass pouring and subsequent heat input which, replaced the heat loss term in the initial model. This new model allowed for changes in flow geometry based on pour rate as well as the ability to increase and decrease flow and stop and restart flow to simulate varying process conditions. A revised COMSOL Multiphysics model was developed to predict temperatures of the glass within DWPF canisters during filling and cooldown. The model simulations and experimental data were in good agreement. The largest temperature deviations were ∼ 40 C for the 87 inch thermocouple location at 3000 minutes and during the initial cool down at the 51 inch location occurring at approximately 600 minutes. Additionally, the model described in this report predicts the general temperature trends during filling and cooling as observed experimentally. The revised model incorporates a heat flow region corresponding to the glass pouring down the centerline of the canister. The geometry of this region is dependent on the flow rate of the glass and can therefore be used to see temperature variations for various pour rates. The equations used for this model were developed by comparing simulation output to experimental data from a single pour rate. Use of the model will predict temperature profiles for other pour rates but the accuracy of the simulations is unknown due to only a single flow rate comparison.

  8. Pitting corrosion on a copper canister; Gropfraetning paa kopparkapsel

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P.; Beverskog, B. [Studsvik Material AB, Nykoeping, (Sweden)

    1996-02-01

    It is demonstrated that normal pitting can occur during oxidizing conditions in the repository. It is also concluded that a new theory for pitting corrosion has to be developed, as the present theory is not in accordance with all practical and experimental observations. A special variant of pitting, based on the growth of sulfide whiskers, is suggested to occur during reducing conditions. However, such a mechanism needs to be demonstrated experimentally. A simple calculational model of canister corrosion was developed based on the results of this study. 69 refs, 3 figs.

  9. The canister durability tests of the in-can type incineration-melting furnace

    International Nuclear Information System (INIS)

    Construction of LEDF (Large equipment dismantling facility) which has the in-can type incineration-melting furnace is planned. The in-can type incineration-melting furnace performs incineration and melting solidification of radioactive waste within the canister made from ceramics, and is characterized by discarding the canister. On the other hand, as for this furnace, the amount of incineration is restrained to canister capacity. Therefore, how to repeat incineration and melting can be considered as a method of increasing the amount of incineration. However, we were anxious about the contact time of the melt and a canister extending, the amount of wear of canister base material increasing, or the heat load (heat cycle) to a canister increasing, and the material intensity of canister base material falling, in order that this method may repeat incineration and melting. Then, the tests used imitation waste, are the conditions which repeat(1,3,10 bathes) the incineration temperature of 1000degC, and the melt temperature of 1500degC, and investigated change of the amount of wear of canister base material and high temperature bend strength. The result is as follows. (1) The amount of wear of canister base material was 0.09 mm/h at the maximum. This result was a sufficiently few value, even if compared with the conventional result (1.0 mm /h). Moreover, the high temperature bend strength of canister base material is about 3 Mpa on an average, and change was not seen before and after the examination to which heat load is applied. (2) These tests showed that the factor which spoils the soundness of a canister was oxidisation degradation of the canister base material by peeling from the base material of Glaze (glass coating material). The portion embrittlement by oxidisation degradation is locally worn down by contact of the melt. (3) Heat-resistant temperature of Glaze is about 1300degC. At the melting operation temperature of 1500degC, and the incineration temperature of

  10. 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).

  11. The P6 truss moves to a payload transport canister

    Science.gov (United States)

    2000-01-01

    In the Space Station Processing Facility, the P6 integrated truss segment is placed in the payload transport canister while workers watch its progress. After being secured in the canister, the truss will be transported to Launch Pad 39B and the payload changeout room. Then it will be moved into Space Shuttle Endeavour's payload bay for mission STS-97. The P6 comprises Solar Array Wing-3 and the Integrated Electronic Assembly, to be installed on the Space Station. The Station's electrical power system will use eight photovoltaic solar arrays, each 112 feet long by 39 feet wide, to convert sunlight to electricity. The solar arrays are mounted on a '''blanket''' that can be folded like an accordion for delivery. Once in orbit, astronauts will deploy the blankets to their full size. Gimbals will be used to rotate the arrays so that they will face the Sun to provide maximum power to the Space Station. The STS-97 launch is scheduled Nov. 30 at 10:06 p.m. EST.

  12. Plutonium Can-In-Canister-Design Basis Event Analysis

    International Nuclear Information System (INIS)

    The purpose of this document is to perform a preliminary design basis event (DBE) analysis of the immobilized plutonium (can-in-canister) waste form to be referred to in this analysis as high level waste/plutonium (HLW/Pu). The objective of the analysis is to determine any preclosure safety impacts of the waste form on the Monitored Geologic Repository (MGR). The scope of this analysis is to determine the offsite dose consequences and associated frequencies of selected DBEs for systems handling disposable canisters that bound all surface and subsurface off-normal events, and to compare these results against regulatory limits. The results of this work are preliminary and are intended to be used to establish a set of preliminary MGR and waste form requirements, to identify mitigation or prevention options that may be required to meet regulatory limits, and to provide input to the Site Recommendation (SR) report. This document is prepared in accordance with the associated development plan (Civilian Radioactive Waste Management System Management and Operating Contractor [CRWMS M and O] 1999e)

  13. 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.

  14. Test plan for K Basin Sludge Canister and Floor Sampling Device

    Energy Technology Data Exchange (ETDEWEB)

    Meling, T.A.

    1995-03-28

    This document provides the test plan and procedure forms for conducting the functional and operational acceptance testing of the K Basin Sludge Canister and Floor Sampling Device(s). These samplers samples sludge off the floor of the 100K Basins and out of 100K fuel storage canisters.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  17. The design analysis of ACP-canister for nuclear waste disposal

    International Nuclear Information System (INIS)

    The design basis, dimensioning and some manufacturing aspects of the Advanced Cold Process Canister (ACPC) for the nuclear waste disposal is summarized in the report. The strength of the canister has been evaluated in normal design load condition and in extreme high hydrostatic pressure load condition possibly caused by ice age (orig.)

  18. Thermal behavior of the CANDU type spent fuel dry-storage concrete canister

    International Nuclear Information System (INIS)

    This paper describes a simple model developed for calculation of the temperature distribution and thermal behavior analysis of the spent fuel dry-storage concrete canister. The model takes into account the relevant heat transfer processes and the cylindrical geometry of the concrete canister. (author)

  19. 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.

  20. HANSF 1.3.2 User's Manual

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  4. Canister cryogenic system for cooling germanium semiconductor detectors in borehole and marine probes

    Science.gov (United States)

    Boynton, G.R.

    1975-01-01

    High resolution intrinsic and lithium-drifted germanium gamma-ray detectors operate at about 77-90 K. A cryostat for borehole and marine applications has been designed that makes use of prefrozen propane canisters. Uses of such canisters simplifies cryostat construction, and the rapid exchange of canisters greatly reduces the time required to restore the detector to full holding-time capability and enhances the safety of a field operation where high-intensity 252Cf or other isotopic sources are used. A holding time of 6 h at 86 K was achieved in the laboratory in a simulated borehole probe in which a canister 3.7 cm diameter by 57 cm long was used. Longer holding times can be achieved by larger volume canisters in marine probes. ?? 1975.

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

    International Nuclear Information System (INIS)

    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% SiO2 and 15% B2O3. Leachabilities of SRP waste glasses are expected to approach 10-8 g/m2-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

  6. Remote Handled WIPP Canisters at Los Alamos National Laboratory Characterized for Retrieval

    International Nuclear Information System (INIS)

    The Los Alamos National Laboratory (LANL) is pursuing retrieval, transportation, and disposal of 16 remote handled transuranic waste canisters stored below ground in shafts since 1994. These canisters were retrievably stored in the shafts to await Nuclear Regulatory Commission certification of the Model Number RH-TRU 72B transportation cask and authorization of the Waste Isolation Pilot Plant (WIPP) to accept the canisters for disposal. Retrieval planning included radiological characterization and visual inspection of the canisters to confirm historical records, verify container integrity, determine proper personnel protection for the retrieval operations, provide radiological dose and exposure rate data for retrieval operations, and to provide exterior radiological contamination data. The radiological characterization and visual inspection of the canisters was performed in May 2006. The effort required the development of remote techniques and equipment due to the potential for personnel exposure to radiological doses approaching 300 R/hr. Innovations included the use of two nested 1.5 meter (m) (5-feet [ft]) long concrete culvert pipes (1.1-m [42 inch (in.)] and 1.5-m [60-in] diameter, respectively) as radiological shielding and collapsible electrostatic dusting wands to collect radiological swipe samples from the annular space between the canister and shaft wall. Visual inspection indicated that the canisters are in good condition with little or no rust, the welded seams are intact, and ten of the canisters include hydrogen gas sampling equipment on the pintle that will have to be removed prior to retrieval. The visual inspection also provided six canister identification numbers that matched historical storage records. The exterior radiological data indicated alpha and beta contamination below LANL release criteria and radiological dose and exposure rates lower than expected based upon historical data and modeling of the canister contents. (authors)

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

    International Nuclear Information System (INIS)

    '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 bentonite is

  8. 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

  9. Lattice specific heat for the RMIn5 (R=Gd, La, Y; M=Co, Rh) compounds: Non-magnetic contribution subtraction

    Science.gov (United States)

    Facio, Jorge I.; Betancourth, D.; Cejas Bolecek, N. R.; Jorge, G. A.; Pedrazzini, Pablo; Correa, V. F.; Cornaglia, Pablo S.; Vildosola, V.; García, D. J.

    2016-06-01

    We analyze theoretically a common experimental process used to obtain the magnetic contribution to the specific heat of a given magnetic material. In the procedure, the specific heat of a non-magnetic analog is measured and used to subtract the non-magnetic contributions, which are generally dominated by the lattice degrees of freedom in a wide range of temperatures. We calculate the lattice contribution to the specific heat for the magnetic compounds GdMIn5 (M=Co, Rh) and for the non-magnetic YMIn5 and LaMIn5 (M=Co, Rh), using density functional theory based methods. We find that the best non-magnetic analog for the subtraction depends on the magnetic material and on the range of temperatures. While the phonon specific heat contribution of YRhIn5 is an excellent approximation to the one of GdCoIn5 in the full temperature range, for GdRhIn5 we find a better agreement with LaCoIn5, in both cases, as a result of an optimum compensation effect between masses and volumes. We present measurements of the specific heat of the compounds GdMIn5 (M=Co, Rh) up to room temperature where it surpasses the value expected from the Dulong-Petit law. We obtain a good agreement between theory and experiment when we include anharmonic effects in the calculations.

  10. Reliability in sealing of canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    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

  11. 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

  12. Critical Issues for Long-Term Nuclear Waste Canister Safety: How 'Good' is 'Good Enough?'

    International Nuclear Information System (INIS)

    The long-term performance of KBS-3 canisters for geologic disposal of spent nuclear fuel will depend upon a number of critical issues. This summary provides an overview of these critical issues, which include near-field environmental conditions, metallurgical composition, fabrication history, long-term performance, and the acceptable margin or 'factor of safety' for this performance. The impact of these factors on the mechanical integrity of KBS-3 canisters is also addressed. The KBS-3 canister design was developed to withstand the environmental conditions predicted to occur following the emplacement of the canisters in Bentonite-filled boreholes (or drifts) in a saturated granite repository horizon. This emplacement scenario was conceived to utilize the advantageous effect of Bentonite swelling, which occurs as the repository re-saturates following final closure. Critical issues that will impact the mechanical integrity of the KBS-3 canisters include potential variation in the water composition (fresh vs. saline), the uniformity of the re-saturation of the Bentonite (and the subsequent strains that will be induced on the canisters), the plastic deformation and creep deformation of the copper, outer barrier under 'normal' conditions, and the potential, significant mechanical deformations that may result from seismically induced canister shear. Another set of parameters that has the potential to significantly impact the mechanical integrity of KBS-3 canisters is the metallurgical composition of the copper, outer barrier and the composition and microstructure of this barrier at the final closure seal. Current KBS-3 design plans call for the use of high-purity copper that is seal with either an electron beam weld or a friction stir weld. The methods of fabrication and inspection for both the base metal of the canister and the closure seal will provide the opportunity for undetected 'flaws' that have the potential to compromise the mechanical integrity of the canister

  13. Uncertainty evaluation in radon concentration measurement using charcoal canister

    International Nuclear Information System (INIS)

    Active charcoal detectors are used for testing the concentration of radon in dwellings. The method of measurement is based on radon adsorption on coal and measurement of gamma radiation of radon daughters. The contributions to the final measurement uncertainty are identified, based on the equation for radon activity concentration calculation. Different methods for setting the region of interest for gamma spectrometry of canisters were discussed and evaluated. The obtained radon activity concentration and uncertainties do not depend on peak area determination method. - Highlights: • Measurement uncertainty budget for radon activity concentration established. • Three different methods for ROI selection are used and compared. • Recommend to use one continuous ROI, less sensitive to gamma spectrometry system instabilities

  14. Acoustic monitoring techniques for corrosion degradation in cemented waste canisters

    International Nuclear Information System (INIS)

    This report describes work carried out to investigate acoustic emission as a monitor of corrosion and degradation of wasteforms where the waste is potentially reactive metal. Electronic monitoring equipment has been designed, built and tested to allow long-term monitoring of a number of waste packages simultaneously. Acoustic monitoring experiments were made on a range of 1 litre cemented Magnox and aluminium samples cast into canisters comparing the acoustic events with hydrogen gas evolution rates and electrochemical corrosion rates. The attenuation of the acoustic signals by the cement grout under a range of conditions has been studied to determine the volume of wasteform that can be satisfactorily monitored by one transducer. The final phase of the programme monitored the acoustic events from full size (200 litre) cemented, inactive, simulated aluminium swarf wastepackages prepared at the AEA waste cementation plant at Winfrith. (Author)

  15. 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.

  16. Analysis of sludge from Hanford K East Basin canisters

    International Nuclear Information System (INIS)

    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

  17. Stardust is revealed after a protective canister is removed

    Science.gov (United States)

    1999-01-01

    At Launch Pad 17-A, Cape Canaveral Air Station, workers watch as the protective canister is lifted from the Stardust spacecraft. Preparations continue for liftoff of the Boeing Delta II rocket carrying Stardust on Feb. 6. Stardust is destined for a close encounter with the comet Wild 2 in January 2004. Using a silicon- based substance called aerogel, Stardust will capture comet particles flying off the nucleus of the comet. The spacecraft also will bring back samples of interstellar dust. These materials consist of ancient pre-solar interstellar grains and other remnants left over from the formation of the solar system. Scientists expect their analysis to provide important insights into the evolution of the sun and planets and possibly into the origin of life itself. The collected samples will return to Earth in a sample return capsule to be jettisoned as Stardust swings by Earth in January 2006.

  18. Development of remote-operating welding system of canister cap

    International Nuclear Information System (INIS)

    The authors have developed a remote-operating welding system for mock-up test facilities of vitrification process of high level radio-active waste of nuclear fuel. This system enables cap sealing welding of canister to accomodate a vitrified waste. Supposing the operation is conducted under high level radio-active environment, the system has been considered to be well handled remotely by adopting guide-pin connection of the welding head, and also developing the automatic electrode exchanger, detecting method of work piece set location by means of the electrode itself as a sensor, slipping joint of power cable (work piece side) and shielding gas quality checking method by measuring an arc voltage changes. To ensure high quality welding, welding conditions were fully examined and established according to temperatures of the work piece before welding. (author)

  19. Procedural development for nuclear waste canister impact testing

    International Nuclear Information System (INIS)

    Double containment requirements for transporting nuclear waste in glass form are costly and may not be necessary for some waste forms. To allow single containment, a procedure for examining particle size distribution and the amount of respirable particles generated under accident conditions was needed. A statistically designed experiment was conducted to examine the effects of glass temperature, fill rate and canister drop orientation upon the amount of sub-ten micron particles generated under simulated accident conditions. Measuring such small particles is somewhat inaccurate because of material loss in handling. By assuming a lognormal particle size distribution, the amount of sub-ten micron particles was estimated from the results for the larger measurable particles. Analyses revealed no temperature or fill rate effect but indicated that the amount of respirable particles is affected by drop orientation. This led to identification of a worst case drop orientation to be used in qualification testing. 4 refs., 2 figs

  20. STS-100 MPLM Raffaello is moved to the payload canister

    Science.gov (United States)

    2001-01-01

    KENNEDY SPACE CENTER, Fla. - Workers inside the payload canister wait for the Multi-Purpose Logistics Module Raffaello to be lowered inside. It joins the Canadian robotic arm, SSRMS, already in place. Both elements are part of the payload on mission STS- 100 to the International Space Station. Raffaello carries six system racks and two storage racks for the U.S. Lab. The arm has seven motorized joints and is capable of handling large payloads and assisting with docking the Space Shuttle. The SSRMS is self- relocatable with a Latching End Effector so it can be attached to complementary ports spread throughout the Station'''s exterior surfaces. Launch of STS-100 is scheduled for April 19, 2001 at 2:41 p.m. EDT from Launch Pad 39A.

  1. Local structures in mixed LixFe1-yMyPO4 (M=Co, Ni) electrode materials

    Science.gov (United States)

    Jalkanen, K.; Lindén, J.; Karppinen, M.

    2015-10-01

    We employ 57Fe Mössbauer spectroscopy as a local tool to probe electrical environments of Fe2+ and Fe3+ at different lithiation (x) and cation-substitution (y) levels in LixFe1-yMyPO4/C (M=Co, Ni) Li-ion battery electrode materials. Upon delithiation the local environment of Fe3+ remains unaffected for the parent y=0 system due to the LiFePO4/FePO4 phase separation, whereas for y>0 changes in the electrical environment are seen for Fe3+. When the Fe2+/Fe3+ redox couple is partially-delithiated, a decreasing quadrupole splitting value is observed for Fe3+ with increasing y, implying a more symmetric electrical environment. The increasing concentration of the Co2+/Ni2+ substituent introduces increasing amounts of Li atoms in the Fe3+-containing phase, and these nearest-neighbor Li atoms are suspected to cause the changes seen in the local environment of Fe3+.

  2. The influence of SRT on phosphorus removal and sludge characteristics in the HA-A/A-MCO sludge reduction process

    Science.gov (United States)

    Zuo, N.; Ji, F. Y.

    2013-02-01

    By researching the influence of sludge age (SRT) on phosphorous removal and sludge characteristics in the HA-A/A-MCO (hydrolysis-acidification-anaerobic/anoxic-multistep continuous oxic tank) process, which has the effect of simultaneous phosphorous and nitrogen removal and sludge reduction, it is found that extended SRT is helpful for improving the ability of anaerobic phosphorous release and chemical recovery of phosphate, but the hosphorous removal efficiency is not affected. Extended SRT causes the system to have even more active sludge; it can also lead to the system having a powerful ability of biochemical reaction by using superiority of concentration. Meanwhile, extended SRT can still reduce sludge yield. Extended SRT cannot make soluble metabolic product (SMP) accumulate in the reactor, so that the pollutant removal power is reduced; it also cannot affect the activity of the sludge. However, extended SRT is able to make the coagulation of the sludge hard, and cause the sludge volume index value increase, but cannot cause sludge bulking.

  3. Incommensurate and ferrimagnetic phases in U(Ni,M)2Si2 with minor M=Co or Cu

    International Nuclear Information System (INIS)

    UNi2Si2 orders magnetically at 124±1 K in an incommensurate (IC) phase, undergoes transition at 103±1 K to AF-I (+-+-) phase, and then another transition at 53±1 K to a ferrimagnetic (++-) phase. A.c.-susceptibility and neutron-diffraction studies of polycrystalline U(Ni,M)2Si2 solid solutions, with minor M=Co and Cu (U(Co1-yNiy)2Si2 with y=0.75, 0.90, and 0.95, and U(Ni1-zCuz)2Si2, with z=0.05, 0.10, and 0.25) confirm an IC phase for y=0.90 below 119±2 K down to 105±2 K and suggest (by a.c.-susceptibility only) IC phases below TN for compositions between y∼0.85 and z∼0.05. The above ferrimagnetic phase is observed at T≤12 K also for compositions between y∼0.93 and z∼0.03. The magnetic phase diagram in the vicinity of UNi2Si2 is redrawn and discussed. (orig.)

  4. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for disposal of spent nuclear fuel from reactors of Teollisuuden Voima Oy (TVO) and Imatran Voima Oy (IVO) are discussed. The canister design is based on the Posiva's concept where solid insert structure is surrounded by the copper mantle. During recent years Outokumpu Copper Products and Posiva have continued their work on development of the copper canisters. Outokumpu Copper Products has also increased capability to manufacture these canisters. In the study the most potential manufacturing methods and their costs are discussed. The cost estimates are based on the assumption that Outokumpu will supply complete copper mantles. At the moment there are at least two commercially available production methods for copper cylinder manufacturing. These routes are based on either hot extrusion of the copper tube or hot rolling, bending and EB-welding of the tube. Trial fabrications has been carried out with both methods for the full size canisters. These trials of the canisters has shown that both the forming from rolled plate and the extrusion are possible methods for fabricating copper canisters on a full scale. (orig.) (26 refs.)

  5. 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.

  6. Equipment for deployment of canisters with spent nuclear fuel and bentonite buffer in horizontal holes

    International Nuclear Information System (INIS)

    The study presents the predesign of equipment for the deployment of canisters in long horizontal holes. The canisters are placed in the centre of the hole and are surrounded by a bentonite buffer. In thE study the canisters are assumed to have a diameter of 1.6 m and a length of 5.9 m, including the hemispherical ends. Their total weight is 60 tonnes. The bentonite buffer after homogenization is 400 mm thick, making a total package diameter of 2.4 m. The deployment system consists of four wagons for handling The canisters and the bentonite blocks. To ensure safe emplacement, every part is installed separately in its final position. This also makes it possible to use small clearances between the canisters and the bentonite blocks and between the blocks and the rock wall. With small clearances, backfilling can be avoided. Another basic design idea is that the wagons are equipped with wheels, which are in direct contact with the rock walls. Thus, rails, which have to be removed as the deployment progresses, are unnecessary. To minimize the time taken for deploying one canister, the wagons are designed so that only three trips from the service area to the deposit area are needed. Due to the radiation in the vicinity of the canisters, the wagons have to be teleoperated

  7. Equipment for deployment of canisters with spent nuclear fuel and bentonite buffer in horisontal holes

    International Nuclear Information System (INIS)

    This study presents the predesign of equipment for the deployment of canisters in long horizontal holes. The canisters are placed in the centre of the hole and are surrounded by a bentonite buffer. In this study the canisters are assumed to have a diameter of 1.6 m and a length of 5.9 m, including the hemispherical ends. Their total weight is 60 tonnes. The bentonite buffer after homogenization is 400 mm thick, making a total package diameter of 2.4 m. The deployment system consists of four wagons for handling the canisters and the bentonite blocks. To ensure safe emplacement, every part is installed separately in its final position. This also makes it possible to use small clearances between the canisters and the bentonite blocks and between the blocks and the rock wall. With small clearances, backfilling can be avoided. Another basic design idea is that the wagons are equipped with wheels, which are in direct contact with the rock walls. Thus, rails, which have to be removed as the deployment progresses, are unnecessary. To minimize the time taken for deploying one canister, the wagons are designed so that only three trips from the service area to the deposit area are needed. Due to the radiation in the vicinity of the canisters, the wagons have to be teleoperated. (au)

  8. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    This paper covers the design, development and testing of the magazines (cylinders containing cans of plutonium-ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed robotic arm that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102mm) diameter canister throat

  9. Preliminary design specification for Department of Energy standardized spent nuclear fuel canisters. Volume 2: Rationale document

    International Nuclear Information System (INIS)

    This document (Volume 2) is a companion document to a preliminary design specification for the design of canisters to be used during the handling, storage, transportation, and repository disposal of Department of Energy (DOE) spent nuclear fuel (SNF). This document contains no procurement information, such as the number of canisters to be fabricated, explicit timeframes for deliverables, etc. However, this rationale document does provide background information and design philosophy in order to help engineers better understand the established design criteria (contained in Volume 1 respectively) necessary to correctly design and fabricate these DOE SNF canisters

  10. The dry storage of used fuel in concrete canisters in Canada

    International Nuclear Information System (INIS)

    The Whiteshell Nuclear Research Establishment (WNRE) initiated a program for dry storage of used CANDU fuel in concrete canisters in 1975. Over the past decade, 17 Mg of fuel have been placed in concrete canister storage at WNRE. In 1985, the WNRE concrete canister design was used for the first time commercially for the interim, on-site storage of 67 Mg of fuel from the Gentilly-1 power reactor at the Gentilly site in the Province of Quebec. This report describes various aspects of this interim storage method. The discussion includes the concept, applications, overall operating experience, licensing aspects, and quality assurance standards and their development

  11. Status and use of the Rocky Flats Environmental Technology Site Pipe Overpack Container for TRU waste storage and shipments

    International Nuclear Information System (INIS)

    The Pipe Overpack Container was designed to optimize shipments of high plutonium content transuranic waste from Rocky Flats Environmental Technology Site (RFETS) to Waste Isolation Pilot Plant (WIPP). The container was approved for use in the TRUPACT-II shipping container by the Nuclear Regulatory Commission in February 1997. The container optimizes shipments to WIPP by increasing the TRUPACT-II criticality limit from 325 fissile grams equivalent (FGE) to 2,800 FGE and provides additional shielding for handling wastes with high americium-241 (Am-241) content. The container was subsequently evaluated and approved for storage of highly dispersible TRU wastes and residues at RFETS. Thermal evaluation of the container shows that the container will mitigate the impact of a worst case thermal event from reactive or potentially pyrophoric materials. These materials contain hazards postulated by the Defense Nuclear Facilities Safety Board for interim storage. Packaging these reactive or potentially pyrophoric residues in the container without stabilizing the materials is under consideration at RFETS. The design, testing, and evaluations used in the approvals, and the current status of the container usage, will be discussed

  12. DISPOSAL OF TRU WASTE FROM THE PLUTONIUM FINISHING PLANT IN PIPE OVERPACK CONTAINERS TO WIPP INCLUDING NEW SECURITY REQUIREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Hopkins, A.M.; Sutter, C.; Hulse, G.; Teal, J.

    2003-02-27

    The Department of Energy is responsible for the safe management and cleanup of the DOE complex. As part of the cleanup and closure of the Plutonium Finishing Plant (PFP) located on the Hanford site, the nuclear material inventory was reviewed to determine the appropriate disposition path. Based on the nuclear material characteristics, the material was designated for stabilization and packaging for long term storage and transfer to the Savannah River Site or, a decision for discard was made. The discarded material was designated as waste material and slated for disposal to the Waste Isolation Pilot Plant (WIPP). Prior to preparing any residue wastes for disposal at the WIPP, several major activities need to be completed. As detailed a processing history as possible of the material including origin of the waste must be researched and documented. A technical basis for termination of safeguards on the material must be prepared and approved. Utilizing process knowledge and processing history, the material must be characterized, sampling requirements determined, acceptable knowledge package and waste designation completed prior to disposal. All of these activities involve several organizations including the contractor, DOE, state representatives and other regulators such as EPA. At PFP, a process has been developed for meeting the many, varied requirements and successfully used to prepare several residue waste streams including Rocky Flats incinerator ash, Hanford incinerator ash and Sand, Slag and Crucible (SS&C) material for disposal. These waste residues are packed into Pipe Overpack Containers for shipment to the WIPP.

  13. High-level waste canister storage final design, installation, and testing. Topical report

    International Nuclear Information System (INIS)

    This report is a description of the West Valley Demonstration Project's radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project's vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access

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

    International Nuclear Information System (INIS)

    The purpose of this report is to provide an evaluation of the Canister Storage Building (CSB) design criteria, the design's compliance with the applicable criteria, and the basis for authorization to proceed with construction of the CSB

  15. Canister design concepts for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  16. Evaluation of radiation shielding performance of disposal canister storing PWR spent fuels

    International Nuclear Information System (INIS)

    Radiation shielding is an important factor in designing disposal canister containing spent nuclear fuel(SNF), because intensity for photon and neutron in SNF assembly after 40 year cooling is still high, ∼1015 photons/sec and ∼108 neutrons/sec, respectively. Radiation escaping from the disposal canister emplaced in repository may cause radiolysis and form oxidizing chemical species. This may result in corrosion of canister itself to proceed. Personnel exposure is also important concern. If shielding performance of canister can reduce radiation level to 1mRem/hr, human access without a control on duration and frequency of exposure may be possible. This provides the benefit of more direct human control of waste packages handling and emplacement operations. In this paper, the radiation shielding performance was evaluated based on current reference disposal system to check absorbed dose for radiolysis, and exposure dose for radiation protection

  17. High-level waste canister storage final design, installation, and testing. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

  18. Demonstration of a Solution Film Leak Test Technique and Equipment for the S00645 Canister Closure

    International Nuclear Information System (INIS)

    The purpose of this effort was to demonstrate that the SFT technique, when adapted to a DWPF canister nozzle, is capable of detecting leaks not meeting the Waste Acceptance Product Specifications (WAPS) acceptance criterion

  19. Demonstration of a Solution Film Leak Test Technique and Equipment for the S00645 Canister Closure

    Energy Technology Data Exchange (ETDEWEB)

    Cannell, G.R.

    1999-10-07

    The purpose of this effort was to demonstrate that the SFT technique, when adapted to a DWPF canister nozzle, is capable of detecting leaks not meeting the Waste Acceptance Product Specifications (WAPS) acceptance criterion.

  20. Multiple-canister flow and transport code in 2-dimensional space. MCFT2D: user's manual

    International Nuclear Information System (INIS)

    A two-dimensional numerical code, MCFT2D (Multiple-Canister Flow and Transport code in 2-Dimensional space), has been developed for groundwater flow and radionuclide transport analyses in a water-saturated high-level radioactive waste (HLW) repository with multiple canisters. A multiple-canister configuration and a non-uniform flow field of the host rock are incorporated in the MCFT2D code. Effects of heterogeneous flow field of the host rock on migration of nuclides can be investigated using MCFT2D. The MCFT2D enables to take into account the various degrees of the dependency of canister configuration for nuclide migration in a water-saturated HLW repository, while the dependency was assumed to be either independent or perfectly dependent in previous studies. This report presents features of the MCFT2D code, numerical simulation using MCFT2D code, and graphical representation of the numerical results. (author)

  1. Coupled transport/reaction modelling of copper canister corrosion aided by microbial processes

    International Nuclear Information System (INIS)

    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 sulphide 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 intruding groundwater, 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 is included in the model. The local depth of copper canister corrosion is calculated by the model. (orig.)

  2. Coupled transport/reaction modelling of copper canister corrosion aided by microbial processes

    Energy Technology Data Exchange (ETDEWEB)

    Liu Jinsong; Neretnieks, I. [Dept. of Chemical Engineering and Technology, Royal Inst. of Tech., Stockholm (Sweden)

    2004-07-01

    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 sulphide 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 intruding groundwater, 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 is included in the model. The local depth of copper canister corrosion is calculated by the model. (orig.)

  3. Multi-dimensional neutronics analysis of the 'canister blanket' for NET

    International Nuclear Information System (INIS)

    At KfK a design of a helium-cooled ceramic breeder blanket, called 'canister blanket', has been developed for the NET fusion test reactor. In this report a detailed neutronic analysis of the 'canister blanket', based on one-, two- and three-dimensional Monte-Carlo calculations in the NET-III double null configuration, is presented. The main object refers to the three-dimensional analysis of a complete sector of the NET-reactor containing the 'canister blanket'. This concerns the poloidal distribution of the neutron wall load and the neutron fluxes at the first wall, the spatial distribution of the power density, the total power production and global effects on the tritium breeding ratio. It is shown that, in case of the 'canister blanket', a global tritium breeding ratio beyond 1.0 seems to be feasible for NET. (orig.)

  4. Process and machinery description of equipment for deposition of canisters in horizontal deposition holes

    International Nuclear Information System (INIS)

    In this report are presented seventeen methods to deposit canisters with spent nuclear fuel in horizontal holes, one canister per hole, in the KBS-3 system. They have been developed successively, after an analysis of weak points and strong points in previously described methods. In conformance with the guidelines for Project JADE, two choices of system have been considered during the development work. One choice is whether the canister should be provided with a tubular radiation shield or not during transport in the secondary tunnels. Another choice is whether canister and bentonite buffer should be deposited at different occasions, but shortly after each other ('in parts') or together in a single package ('in a package'). The basic technical problem is placing heavy objects, the canister and the buffer components, in an horizontal hole which is 8 m long. For depositing of bentonite buffer and canister 'in parts', the use of a guiding pipe has been studied to reduce the impact of a sliding canister on the bentonite rings. For depositing 'in a package', three alternative techniques have been studied: a loading laddle that is rotated, a fork carriage and rails. Development has been aimed at avoiding the use of a guiding pipe and at reducing the cross section area of the secondary tunnel. A failure mode and effect analysis has been performed for three of the methods in order to provide a basis for a decision whether to use a tubular radiation shield around the canister during transport and handling in the secondary tunnels. SKB has subsequently decided, partly on this basis, that the canisters should be placed in radiation shields. The development work reported here has not yet yielded a definitive method for placing canisters in horizontal holes. It is recommended that in the continued work: canister and bentonite buffer are deposited in a hole at the same time, as a package; methods involving a minimum number of movements in the tunnel are preferred and that

  5. Near-Field Mechanical Analysis of Radioactive Waste Canister in Deep Repository

    International Nuclear Information System (INIS)

    The spent nuclear fuel and the radioactive materials formed during the operation of the Swedish nuclear power plants will be enclosed into tight metal canisters. These canisters will then be placed in large disposal boreholes drilled into the floor of the repository tunnels. Bentonite blocks will be placed to fill the space between the canisters and the boreholes. The main purpose with the bentonite is to provide a hydrological barrier. In general the types of analysis required to study the behavior of the canister and the buffer material shall account for mechanical, hydraulic, thermal and chemical effects. In this study, only near field mechanical behavior is investigated. Preliminary analyses are made based on simplified assumptions and on some simple two-dimensional finite element solutions. As a results of the preliminary analysis, limited tectonic movements in the bedrock and unfavorable local swelling are studied and modeled by the finite element code ABAQUS using tree-dimensional models. The bentonite is modeled using two different material models, Mohr-Coulomb and Drucker-Prager, while the canister materials are modeled using a Drucker-Prager material model. A certain form of sensitivity analysis for parameters has also been carried out. The analyses of uneven swelling of the bentonite did not give any plastic strains in the canister. Local swelling is therefore not a threat against the canister. This load case is not the critical one. The results from the analyses of movements in the bedrock show that, as a consequence of large deviatoric stresses, plastic strains appear locally in the canister. However, the material properties for the materials in the canister show that the size of the deviatoric stresses is less than half on the failure stress. Thus, there seems to be no risk for local or total failure of the canister in case of movements in the bedrock. The conclusion from the finite element analyses is that the design of the nuclear waste canister

  6. Process and machinery description of equipment for deposition of canisters in medium-long deposition holes

    International Nuclear Information System (INIS)

    In this report twelve methods are presented to deposit a canister with spent nuclear fuel in a horizontal hole, several canisters per hole (MLH). These methods are part of the KBS-3 system. They have been developed successively, after an analysis of weak points and strong points in previously described methods. In conformance with the guidelines for Project JADE, a choices of system has been considered during the development work. This is whether canister and bentonite buffer should be deposited 'in parts', i.e. at different occasions, but shortly after each other or 'in a package', i.e. together in a single package. The other choice in the guidelines for the JADE project, whether the canister should be placed in a radiation shield or not during transport in the secondary tunnels, was not relevant to MLR. The basic technical problem is depositing heavy objects, the canister and the buffer components, in an horizontal hole which is approximately 200 m deep. Two methods for depositing of the bentonite barrier and the canisters in separate processes have been studied. For depositing of the bentonite barrier and the canister 'in a package', four alternative techniques have been studied: a metallic sleeve around the package, a loading scoop that is rotated, a fork carriage and rails. The repeated transports in a hole, a consequence of depositing several canisters in the same hole, could lead to the rock being crushed. The mutual impact of machines, load and rock wall has therefore been particularly considered. In several methods, the use of a gangway has been proposed (steel plates or layer of ice). A failure mode and effect analysis has been performed for one of the twelve methods. When comparing with a method to deposit one canister per hole using the same technique, the need for equipment and resources is far larger for this MLH method if incidents should occur during depositing. The development work reported here has not yet yielded a definitive method for placing

  7. Local structures in mixed LixFe1−yMyPO4 (M=Co, Ni) electrode materials

    International Nuclear Information System (INIS)

    We employ 57Fe Mössbauer spectroscopy as a local tool to probe electrical environments of Fe2+ and Fe3+ at different lithiation (x) and cation-substitution (y) levels in LixFe1−yMyPO4/C (M=Co, Ni) Li-ion battery electrode materials. Upon delithiation the local environment of Fe3+ remains unaffected for the parent y=0 system due to the LiFePO4/FePO4 phase separation, whereas for y>0 changes in the electrical environment are seen for Fe3+. When the Fe2+/Fe3+ redox couple is partially-delithiated, a decreasing quadrupole splitting value is observed for Fe3+ with increasing y, implying a more symmetric electrical environment. The increasing concentration of the Co2+/Ni2+ substituent introduces increasing amounts of Li atoms in the Fe3+-containing phase, and these nearest-neighbor Li atoms are suspected to cause the changes seen in the local environment of Fe3+. - Graphical abstract: Local environment of iron in LixFe1−y(Co/Ni)yPO4 is studied by 57Fe Mössbauer spectroscopy at different lithiation (x) and cation-substitution (y) levels. - Highlights: • Local Fe environment in LixFe1−y(Co/Ni)yPO4 is studied by 57Fe Mössbauer spectroscopy. • Co/Ni-for-Fe substitution results in a more symmetric electrical environment for Fe3+. • Due to presence of Co2+/Ni2+, Li atoms are introduced into the Fe3+-containing phase. • These nearest-neighbor Li atoms are suggested to change the local Fe3+ environment

  8. Data compilation report: Gas and liquid samples from K West Basin fuel storage canisters

    International Nuclear Information System (INIS)

    Forty-one gas and liquid samples were taken from spent fuel storage canisters in the K West Basin during a March 1995 sampling campaign. (Spent fuel from the N Reactor is stored in sealed canisters at the bottom of the K West Basin.) A description of the sampling process, gamma energy analysis data, and quantitative gas mass spectroscopy data are documented. This documentation does not include data analysis

  9. Gas and liquid sampling for closed canisters in K-West basins - functional design criteria

    International Nuclear Information System (INIS)

    The purpose of this document is to provide functions and requirements for the design and fabrication of equipment for sampling closed canisters in the K-West basin. The samples will be used to help determine the state of the fuel elements in closed canisters. The characterization information obtained will support evaluation and development of processes required for safe storage and disposition of Spent Nuclear Fuel (SNF) materials

  10. Application of the TEMPEST computer code to canister-filling heat transfer problems

    International Nuclear Information System (INIS)

    Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch filling mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs

  11. Data quality objectives for gas and liquid samples from sealed K Basin canisters

    International Nuclear Information System (INIS)

    Data Quality Objectives (DQOS) for gas and liquid sampling from the sealed canisters in K West Basin have been developed and are presented in this document. The scope of this document is limited primarily to the initial sampling effort. This sampling campaign either supports the selection of canisters to provide fuel for hotcell examinations, supports the demonstration of sampling equipment capabilities or provides an initial assessment of gas/liquid chemistry for comparison to the results of fuel element hotcell examinations. No sampling of canisters has occurred since 1983. It is proposed here that samples of gas and water be analyzed for constituents such as cesium, fission gases, and hydrogen which are markers for corrosion of uranium in a water environment. These data will allow an assessment of the risks involved when particular canisters are opened to retrieve fuel. This sampling campaign will also ensure that canisters with some failed fuel elements are included in the population that is opened for retrieval of fuel for hotcell examinations. Additionally, valuable correlations between the macroscopic visible condition of fuel, hotcell examinations, and the gases generated in canisters will be possible. The analysis of other chemical species in the gas and liquid will allow assessments of the performance of the previously added corrosion inhibitor and possibly assessments of radiolysis. Sampling of canisters will be performed with equipment that opens the valves in the canister lid and draws a 15 ml sample of either gas or water. This work will most likely be performed in one of the pits-associated with the K West Basin

  12. An assessment of canister needs for defueling the TMI-2 core

    International Nuclear Information System (INIS)

    It is projected that the TMI-2 Cleanup Program can be completed with a total of 355 canisters (272 fuel, 75 filter, and 8 k/o canisters). This is within the 360 canister space allocation at the INEL. There is a sufficient number and mix of available canisters on-site to meet the outstanding requirements. As of May 1989, the shipment campaign has included 18 rail shipments, with a total of 259 canisters. It is estimated that an additional five rail shipments of three casks (21 canisters) each will be required to complete the program. The achievements of the shipment campaign, the challenges that have been presented, and the reasons for its success can be outlined as follows: very few reactors have ever had to undertake a fuel shipment program paralleled to the magnitude of the TMI-2 program; the cleanup project faced a task of transporting an entire damaged reactor core from TMI-2 to the INEL; this shipment campaign may one day become a blueprint for future shipments of spent fuel by other utilities; the transport system essentially consists of three major subsystems: the casks, the cask support systems, and the shielded dry fuel transfer system, the program successfully worked out the interactions and operation of these subsystems; to date, the shipment program has compiled an impressive record of safe, on-time, and essentially trouble-free performance

  13. 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)

  14. The Meaning of the Sampling of the ZPPR Canisters And Proposed New Surveillance Operating Instructions

    Energy Technology Data Exchange (ETDEWEB)

    Charles W. Solbrig

    2007-01-01

    Analysis of the sample data taken from the ZPPR canisters containing Uranium plate fuel indicates that (as of February 2004) hydriding could be occurring in 35 of them. Since there appears to be no way of determining that a getter is functional, the getters in all the canisters should be replaced now (unless canister residence time can be determined) to prevent further hydriding. In addition, the surveillance procedure should be modified. Canisters to be inspected should be selected sequentially, 12 each quarter resulting in all being opened once every five years. Three of the 12 should be sampled and results reported before opening any of the canisters. Water vapor and pressure should be measured as well as the current hydrogen, oxygen, and nitrogen. Then all 12 canisters should be opened for physical evaluation of the plate conditions and correlation with the sample measurements. The getters should be replaced at each inspection ensuring that no getter is used more than five years. The data should be analyzed each year and a conclusion made on the adequacy of the surveillance procedure and modifications made if it is inadequate.

  15. Effects of annular air gaps surrounding an emplaced nuclear waste canister in deep geologic storage

    International Nuclear Information System (INIS)

    Annular air spaces surrounding an emplaced nuclear waste canister in deep geologic storage will have significant effects on the long-term performance of the waste form. Addressed specifically in this analysis is the influence of a gap on the thermal response of the waste package. Three dimensional numerical modeling predicts temperature effects for a series of parameter variations, including the influence of gap size, surface emissivities, initial thermal power generation of the canister, and the presence/absence of a sleeve. Particular emphasis is placed on determining the effects these variables have on the canister surface temperature. We have identified critical gap sizes at which the peak transient temperature occurs when gap widths are varied for a range of power levels. It is also shown that high emissivities for the heat exchanging surfaces are desirable, while that of the canister surface has the greatest influence. Gap effects are more pronounced, and therefore more effort should be devoted to optimal design, in situations where the absolute temperature of the near field medium is high. This occurs for higher power level emplacements and in geomedia with low thermal conductivities. Finally, loosely inserting a sleeve in the borehole effectively creates two gaps and drastically raises the canister peak temperature. It is possible to use these results in the design of an optimum package configuration which will maintain the canister at acceptable temperature levels. A discussion is provided which relates these findings to NRC regulatory considerations

  16. Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

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

    International Nuclear Information System (INIS)

    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)

  18. 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).

  19. Corrosion of iron: A study for radioactive waste canisters

    Science.gov (United States)

    Lagha, S. Ben; Crusset, D.; Mabille, I.; Tran, M.; Bernard, M. C.; Sutter, E.

    2007-05-01

    The purpose of this study is to examine the risks of atmospheric corrosion of steel waste canisters following their deep geological disposal in the temperature range from 303 to 363 K. The work was performed using iron samples deposited as thin films on a quartz crystal microbalance (QCM) and disposed in a climatic chamber. The experiments showed that, in the temperature under study (298-363 K), the mass increase due to the formation of oxide/hydroxide rose sharply above 70% RH, as is commonly observed at room temperatures, indicating that the phenomenon remains electrochemical in nature. Ex situ Raman spectrometric analyses indicate the formation of magnetite, maghemite and oxyhydroxides species in the 298-363 K temperature range, and for oxygen contents above 1 vol.%, whereas only Fe3O4 and γ-Fe2O3 are detected at 363 K. In this work, the kinetics of the rust growth is discussed, on the bases of the rate of mass increase and of the composition of the rust, as a function of the climatic parameters and the oxygen content of the atmosphere.

  20. Design basis for the copper canister. Stage one

    International Nuclear Information System (INIS)

    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

  1. Three-Dimensional Thermal Modeling of Dry Spent Nuclear Fuel Storage Canisters

    International Nuclear Information System (INIS)

    One of the interim storage configurations being considered for aluminum-clad foreign research reactor fuel, such as the Material and Testing Reactor (MTR) design, is a dry storage facility. To support design studies of storage options, a computational and experimental program was conducted at the Savannah River Site (SRS). The objective was to develop computational fluid dynamics (CFD) conjugate models which would be benchmarked using data obtained from a full scale heat transfer experiment conducted in the SRS Experimental Thermal Fluids Laboratory. The current work describes the modeling approach and presents comparison of computational results with experimental data. The experimental set up consists of an instrumented fuel canister 16 inches in diameter and 36 inches in height.The canister contains a sealed fuel can which is designed to store four fuel assemblies. The fuel assembly heat generation is simulated by an imbeded electrical heater. Each fuel assembly is separated from the others by a stainless steel grid and the assemblies can communicate thermal-hydraulically only through narrow slot holes located at the top and bottom of the assembly. The flow within an enclosed canister is a buoyancy-induced motion resulting from body force acting on density gradients which arise from fluid temperature gradients. The canister is filled with helium or nitrogen gas. The heated canister is surrounded by five unheated dummy canisters and is located inside a wind tunnel. During the test, data are obtained for the radial and axial heat flux/temperature profiles inside the canister, air velocity outside the canister, and ambient air temperature. CFD approach has been used to model the three-dimensional convective velocity and temperature distributions within a single dry storage canister of MTR fuel elements.The final analysis was made for the cases with internal heat source of 85 to 138 watts per MTR fuel element (equivalent to 22 to 35 kW/m3) using various different

  2. Phased Startup Initiative Phase 3 and 4 Test Procedure (OCRWM)

    International Nuclear Information System (INIS)

    The purpose of this test procedure is to safely operate the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) with specific fuel canisters, and show that canisters containing fuel can be retrieved from the canister queue, decapped in the Canister Decapper, and loaded into the Primary Clean Machine (PCM) for fuel cleaning; and that fuel can be sorted on the Process Table, then loaded back into fuel canisters and relocated in basin storage. An option is included to load selected elements into multi-canister overpack (MCO) Fuel Baskets. Additional Data are collected during this test, beyond that collected during production operations. These data support qualifying the cleaning performance of the PCM, assessing the quantity of scrap generated during the cleaning, and evaluating the impact of fuel retrieval operations on the Basin water quality. The additional data collected primarily consist of weighing fuel and scrap at selected points in the operation, as well as photographing fuel and scrap as it is processed. The time to perform operations is also monitored for comparison with design predictions. Water quality data are collected to establish a baseline to predict the effectiveness of equipment design for control of contamination and visibility during production operation

  3. Rare-earth transition-metal chalcogenides Ln3MGaS7 (Ln=Nd, Sm, Dy, Er; M=Co, Ni) and Ln3MGaSe7 (Ln=Nd, Sm, Gd, Dy, M=Co; Ln=Nd, Gd, Dy, M=Ni)

    International Nuclear Information System (INIS)

    Fifteen new rare-earth transition-metal chalcogenides, Ln3MGaS7 (Ln=Nd, Sm, Dy, Er; M=Co, Ni) and Ln3MGaSe7 (Ln=Nd, Sm, Gd, Dy, M=Co; Ln=Nd, Gd, Dy, M=Ni), have been synthesized by solid state reactions. They are isostructural, adopt Ce3Al1.67S7—related structure type, and crystallize in the non-centrosymmetric hexagonal space group P63. They adopt a three-dimensional framework composed of LnQ7 monocapped trigonal prisms with the interesting [MQ3]4− chains and isolated GaQ4 tetrahedra lying in two sets of channels in the framework. The magnetic susceptibility measurements on Ln3CoGaQ7 (Ln=Dy, Er, Q=S; Ln=Dy, Q=Se) indicate that they are paramagnetic and obey the Curie–Weiss law over the entire experimental temperature, while the magnetic susceptibility of Sm3CoGaSe7 deviates from the Curie–Weiss law as a result of the crystal field splitting. - Graphical abstract: Ln3MGaS7 (Ln=Nd, Sm, Dy, Er; M=Co, Ni) and Ln3MGaSe7 (Ln=Nd, Sm, Gd, Dy, M=Co; Ln=Nd, Gd, Dy, M=Ni) adopt a three-dimensional framework composed of LnQ7 monocapped trigonal prisms with interesting [MQ3]4− chains and isolated GaQ4 tetrahedra lying in two sets of channels in the framework. - Highlights: • New compounds, Ln3MGaQ7 (Ln=rare-earth; M=Co, Ni; Q=S, Se), were synthesized. • They are isostructural and crystallize in the noncentrosymmetric space group P63. • They adopt a three-dimensional framework built by LnQ7 monocapped trigonal prisms. • Ln3CoGaQ7 (Ln=Dy, Er; Q=S, Se) are paramagnetic and obey the Curie–Weiss law. • The magnetic susceptibility of Sm3CoGaSe7 deviates from the Curie–Weiss law

  4. 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

  5. Nanomembrane Canister Architectures for the Visualization and Filtration of Oxyanion Toxins with One-Step Processing.

    Science.gov (United States)

    Aboelmagd, Ahmed; El-Safty, Sherif A; Shenashen, Mohamed A; Elshehy, Emad A; Khairy, Mohamed; Sakaic, Masaru; Yamaguchi, Hitoshi

    2015-11-01

    Nanomembrane canister-like architectures were fabricated by using hexagonal mesocylinder-shaped aluminosilica nanotubes (MNTs)-porous anodic alumina (PAA) hybrid nanochannels. The engineering pattern of the MNTs inside a 60 μm-long membrane channel enabled the creation of unique canister-like channel necks and cavities. The open-tubular canister architecture design provides controllable, reproducible, and one-step processing patterns of visual detection and rejection/permeation of oxyanion toxins such as selenite (SeO3(2-)) in aquatic environments (i.e., in ground and river water sources) in the Ibaraki Prefecture of Japan. The decoration of organic ligand moieties such as omega chrome black blue (OCG) into inorganic Al2O3@tubular SiO2/Al2O3 canister membrane channel cavities led to the fabrication of an optical nanomembrane sensor (ONS). The OCG ligand was not leached from the canister as observed in washing, sensing, and recovery assays of selenite anions in solution, which enabled its multiple reuse. The ONS makes a variety of alternate processing analyses of selective quantification, visual detection, rejection/permeation, and recovery of toxic selenite quick and simple without using complex instrumentation. Under optimal conditions, the ONS canister exhibited a high selectivity toward selenite anions relative to other ions and a low-level detection limit of 0.0093 μM. Real analytical data showed that approximately 96% of SeO3(2-) anions can be recovered from aquatic and wastewater samples. The ONS canister holds potential for field recovery applications of toxic selenite anions from water. PMID:26178184

  6. Transporting existing VSC-24 canisters using a risk-based licensing approach

    International Nuclear Information System (INIS)

    The eventual disposition of the spent fuel assemblies loaded in canisters and casks currently designed and licensed only for on-site storage is an industry-wide issue. The canister-specific BUC evaluation approach developed by BFS can be used to license many of these storage canisters and casks for transportation. This will allow these storage canisters and casks to be transported intact to a long-term storage facility or repository, thereby minimizing fuel handling operations, impact on plant operations, and occupational exposure, as well as total infrastructure costs. Application of the proposed canister-specific BUC analysis approach to a preliminary evaluation of the 58 loaded MSBs demonstrates the benefits of this approach. The results of this preliminary evaluation show that a more rigorous analysis based on the known characteristics of the loaded spent fuel, rather than the design-basis fuel parameters, produces significantly lower maximum keff values and can be used to qualify many of the existing loaded storage canisters for transportation. Transportation certification for storage canisters having more reactive spent fuel payloads may require reliance on BUC approaches that are more aggressive than current NRC guidelines allow. Credit may be required for fission- product isotopes that do not have sufficient chemical assay data for benchmarking. In addition, reduced criticality safety margins may be required. For these more-aggressive BUC approaches, a risk assessment should be provided to support the NRC-approval basis. The risk assessment should evaluate the possibility and consequences of an accidental criticality event based upon inaccuracies in the characterization of the spent-fuel payloads

  7. 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

  8. Manufacturing of the canister shells T54 and T55

    International Nuclear Information System (INIS)

    This report constitutes a summary of the manufacturing test of the disposal canister copper shells T54 and T55. The copper billets were manufactured at Luvata Pori Oy, Finland. The hot-forming and machining of the copper shells were made at Vallourec and Mannesmann Tubes, Reisholz mill, Germany. The shells were manufactured with the pierce and draw method. Both of the pipes were manufactured separately in two phases. The first phase consisted of following steps: preheating of the billet, upsetting, piercing and the first draw with mandrel through drawing ring. After cooling down the block is measured and machined in case of excessive eccentricity or surface defects. In the second phase the block is heated up again and expanded and drawn in 6 sequences. In this process the pipe inside dimension is expanded and the length is increased in each step. Before the last, the 6th step, the bottom of the pipe is deformed in a sequence of special processes. During the manufacture of the first pipe, T54, some difficulties were detected with the centralization of the billet before upsetting. For the second manufacture of the T55, an additional steering ring was made and the result was remarkably more coaxial. After the manufacture and non-destructive inspections the shells were cut in pieces and three parts of each shell were taken for destructive testing. The three inspected parts were the bottom plate, a ring from the middle of the cylinder and a ring from the top of the cylinder. The destructive testing was made by Luvata Pori Oy. In spite of some practical difficulties and accidents during the manufacturing process, the results of the examinations showed that both of the test produced copper shells fulfilled all the specified requirements as for soundness (integrity), mechanical properties, chemical composition, dimensions, hardness and grain size. (orig.)

  9. Ruthenium and osmium carbonyl nitrosyl complexes: Matrix infrared spectra and density functional calculations for M(CO)2(NO)2 and M(CO)(NO) (M = Ru, Os)

    International Nuclear Information System (INIS)

    Highlights: ► Laser-ablated ruthenium or osmium atom reactions with CO and NO mixtures in solid argon. ► Metal carbonyl nitrosyls including M(CO)(NO) and 18-electron configuration M(CO)2(NO)2 molecules (M = Ru, Os). ► The observed absorption bands of reaction products are identified by isotopic substitution and DFT calculations. ► The bonding and reaction mechanism are discussed in detail. -- Abstract: Laser-ablated ruthenium or osmium atom reactions with CO and NO mixtures in solid argon produce unsaturated metal carbonyl nitrosyls including M(CO)(NO) and 18-electron configuration M(CO)2(NO)2 molecules (M = Ru, Os). The observed absorption bands of reaction products are identified by isotopic substitution, isotopic ratios and isotopic distributions (13CO, 15NO, and mixtures). DFT (B3LYP and BP86) vibrational fundamental calculations reproduce observed frequencies and isotopic shifts very well. The bonding and reaction mechanism are discussed.

  10. 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

  11. New rock salt-related oxides Li3M2RuO6 (M=Co, Ni): Synthesis, structure, magnetism and electrochemistry

    International Nuclear Information System (INIS)

    We describe the synthesis, crystal structure, magnetic and electrochemical characterization of new rock salt-related oxides of formula, Li3M2RuO6 (M=Co, Ni). The M=Co oxide adopts the LiCoO2 (R-3m) structure, where sheets of LiO6 and (Co2/Ru)O6 octahedra are alternately stacked along the c-direction. The M=Ni oxide also adopts a similar layered structure related to Li2TiO3, where partial mixing of Li and Ni/Ru atoms lowers the symmetry to monoclinic (C2/c). Magnetic susceptibility measurements reveal that in Li3Co2RuO6, the oxidation states of transition metal ions are Co3+ (S=0), Co2+ (S=1/2) and Ru4+ (S=1), all of them in low-spin configuration and at 10 K, the material orders antiferromagnetically. Analogous Li3Ni2RuO6 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 Li3Ni2RuO6 when compared to Li3Co2RuO6. Interestingly high first charge capacities (between ca. 160 and 180 mAh g−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 LiCoO2. - Graphical abstract: Two new rock salt related oxides of formula, Li3M2RuO6, (M=Co, Ni) have been prepared. The M=Co oxide adopts the LiCoO2 (R-3m) structure and the M=Ni oxide adopts a similar layered structure related to Li2TiO3, monoclinic (C2/c), with partial mixing of Li and Ni/Ru atoms. For Li3Co2RuO6, oxidation state for Ru is 4+ and antiferromagnetic (AFM) order is found below 10 K while for the analogous Li3Ni2RuO6 , Ru oxidation state is 5+ and a ferrimagnetic (FM) behavior with a Curie temperature of 100 K is found. Electrochemical studies correlate well with both

  12. Canister transfer in access tunnel. Lay-out, system and operation description

    International Nuclear Information System (INIS)

    In this report the alternative of canister transfer by a vehicle is examined, the principle and the plans are shown in those details that differ from the canister-transfer-throughshaft alternative. In vehicle-transfer alternative the disposal canisters are transferred with a freely steered motor vehicle from ground surface to the repository at level 400 to 500 m below ground surface. The vehicle is a crawler type heavy-load transfer vehicle. The disposal canisters are loaded into the shield cylinder of the vehicle at the encapsulation plant. Canisters are transferred with the vehicle from encapsulation plant to the mouth of the repository ramp, then through the ramp to the repository level underground and finally through central tunnels to the disposal tunnel and disposal hole. Radiation effects of the canister can be detected only in the close vicinity of the vehicle. Transfer route in the site area shall be selected in a way that heavy traffic areas are avoided and the roads used should be even and passable. Underground, the canister transfer proceeds always in the controlled area. The access ramp is declared to be controlled area temporarily in four sections as the transfer proceeds through the ramp. The ventilation is temporarily closed in the controlled area section during canister transfer. To transfer the vehicle from access ramp to the technical rooms of the controlled area of the repository level a construction of a by-pass tunnel is planned. This is made for avoiding disturbance of the simultaneous uncontrolled area operations on the repository level. In two-storey alternative, a by-pass tunnel access is needed also on the lower level of the repository. In case of one-storey repository alternative, the vehicle transfer of the disposal canister does not cause any changes in the order of use of the disposal tunnels or in the organization of controlled and uncontrolled area. In case of two-storey repository, the order of the use of some disposal tunnels is

  13. A review of the possible effects of hydrogen on lifetime of carbon steel nuclear waste canisters

    International Nuclear Information System (INIS)

    In Switzerland, the National Cooperative for the Disposal of Radioactive Waste (Nagra) is responsible for developing an effective method for the safe disposal of vitrified high level waste (HLW) and spent fuel. One of the options for disposal canisters is thick-walled carbon steel. The canisters, which would have a diameter of about 1 m and a length of about 3 m (HLW) or about 5 m (spent fuel), will be embedded in horizontal tunnels and surrounded with bentonite clay. The regulatory requirement for the minimum canister lifetime is 1000 years but demonstration of a minimum lifetime of 10,000 years would be desirable. The pore-water to which the canister will be exposed is of marine origin with about 0.1-0.3 M Cl-. Since hydrogen is generated during the corrosion process, it is necessary to assess the probability of hydrogen assisted cracking modes and to make recommendations to eliminate that probability. To that aim, key reports detailing projections for the local environment and associated corrosion rate of the waste canister have been evaluated with the focus on the implication for the absorbed hydrogen concentration in the steel. Simple calculations of hydrogen diffusion and accumulation in the inner compartment of the sealed canister indicate that a pressure equivalent to that for gas pockets external to the canister (envisaged to be about 10 MPa) may be attained in the proposed exposure time, an important consideration since it is not possible to modify the internal surface of the closure weld. Current ideas on mechanisms of hydrogen assisted cracking are assessed from which it is concluded that the mechanistic understanding and associated models of hydrogen assisted cracking are insufficient to provide a framework for quantitative prediction for this application. The emphasis then was to identify threshold conditions for cracking and to evaluate the likelihood that these may be exceeded over the lifetime of the containment. Based on an analysis of data in the

  14. Analysis of Welding Joint on Handling High Level Waste-Glass Canister

    International Nuclear Information System (INIS)

    The analysis of welding joint of stainless steel austenitic AISI 304 for canister material has been studied. At the handling of waste-glass canister from melter below to interim storage, there is a step of welding of canister lid. Welding quality must be kept in a good condition, in order there is no gas out pass welding pores and canister be able to lift by crane. Two part of stainless steel plate in dimension (200 x 125 x 3) mm was jointed by welding. Welding was conducted by TIG machine with protection gas is argon. Electric current were conducted for welding were 70, 80, 90, 100, 110, 120, 130, and 140 A. Welded plates were cut with dimension according to JIS 3121 standard for tensile strength test. Hardness test in welding zone, HAZ, and plate were conducted by Vickers. Analysis of microstructure by optic microscope. The increasing of electric current at the welding, increasing of tensile strength of welding yields. The best quality welding yields using electric current was 110 A. At the welding with electric current more than 110 A, the electric current influence towards plate quality, so that decreasing of stainless steel plate quality and breaking at the plate. Tensile strength of stainless steel plate welding yields in requirement conditions according to application in canister transportation is 0.24 kg/mm2. (author)

  15. Effects of stabilizers on the heat transfer characteristics of a nuclear waste canister

    International Nuclear Information System (INIS)

    This report summarizes the feasibility and the effectiveness of using stabilizers (internal metal structural components) to augment the heat transfer characteristics of a nuclear waste canister. The problem was modeled as a transient two-dimensional heat transfer in two physical domains - the stabilizer and the wedge (a 30-degree-angle canister segment), which includes the heat-producing spent-fuel rods. This problem is solved by a simultaneous and interrelated numerical investigation of the two domains in cartesian and polar coordinate systems. The numerical investigations were performed for three cases. In the first case, conduction was assumed to be the dominant mechanism for heat transfer. The second case assumed that radiation was the dominant mechanism, and in the third case both radiation and conduction were considered as mechanisms of heat transfer. The results show that for typical conditions in a waste package design, the stabilizers are quite effective in reducing the overall temperature in a waste canister. Furthermore, the results show that increasing the stabilizer thickness over the thickness specified in the present design has a negligible effect on the temperature distribution in the canister. Finally, the presence of the stabilizers was found to shift the location of the peak temperature areas in the waste canister

  16. Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

    International Nuclear Information System (INIS)

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Product Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full scale DWPF canister. The glasses were characterized by X-ray diffraction and scanning electron microscopy to identify the crystalline phases present. The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCT) was used to determine the durability of the heat treated glasses

  17. 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.

  18. Numerical Modelling of Mechanical Integrity of the Copper-Cast Iron Canister. A Literature Review

    International Nuclear Information System (INIS)

    This review article presents a summary of the research works on the numerical modelling of the mechanical integrity of the composite copper-cast iron canisters for the final disposal of Swedish nuclear wastes, conducted by SKB and SKI since 1992. The objective of the review is to evaluate the outstanding issues existing today about the basic design concepts and premises, fundamental issues on processes, properties and parameters considered for the functions and requirements of canisters under the conditions of a deep geological repository. The focus is placed on the adequacy of numerical modelling approaches adopted in regards to the overall mechanical integrity of the canisters, especially the initial state of canisters regarding defects and the consequences of their evolution under external and internal loading mechanisms adopted in the design premises. The emphasis is the stress-strain behaviour and failure/strength, with creep and plasticity involved. Corrosion, although one of the major concerns in the field of canister safety, was not included

  19. Safety evaluation for bolting design of a transportable storage canister of spent nuclear fuels

    International Nuclear Information System (INIS)

    This paper is to perform safety evaluation for bolting design of a transportable storage canister of spent nuclear fuels in a nuclear power plant. To develop the related techniques for inter unit transfer of the spent nuclear fuels, a seamless metal canister design with reopening function is adopted. The canister with bolting flange needs to maintain its seamless and structural integrity under normal operation and postulated accident conditions. For bolting design, the requirements on material and structural strength are completely examined by following ASME Boiler and Pressure Vessel Codes. All calculations in this work are performed by using the commercial finite element analysis software, ANSYS. With different sensitivity analysis results of numerical finite element models, the maximum and minimum operation value of bolting preload torque can be thus obtained. Moreover, during the inter unit transfer and operation of spent nuclear fuels, fatigue of the bolt is addressed and no leakage occurs as the canister keeps closure with lids subject to the accident condition is also verified. The structural functions and safety of a transportable storage canister with new bolting design can be shown.

  20. Certification of VOC canister samplers for use at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    The Waste Isolation Pilot Plant (WIPP) site is designed to demonstrate safe disposal of transuranic (TRU) mixed waste. An air monitoring program has been established at the WIPP site to verify that volatile organic compounds (VOCs) do not migrate out of the disposal unit. In this air monitoring program, modified commercially available dual canister samplers are used to collect air samples for VOC analysis. Sampler certification, sample collection, and sample analysis are performed based on the procedures contained in US Environmental Protection Agency's Compendium Method TO-14. The canister samplers are certified for cleanliness by passing humid zero air through the entire sampling system and collecting a sample in a canister over a 24-hour period. In addition, each canister sampler is certified for target compound recovery efficiency by passing a humid calibration gas standard through the entire sampling system and collecting a sample in a canister over a 24-hour period. In this paper, the authors discuss the techniques developed for meeting the stringent certification requirements of the monitoring program and present data to support the need for these stringent requirements

  1. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

    1996-09-01

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.

  2. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    International Nuclear Information System (INIS)

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments

  3. A preliminary assessment of gas migration from the copper/steel canister

    International Nuclear Information System (INIS)

    A preliminary assessment has been carried out of the consequences of hydrogen gas generation in the copper/steel canister, a new concept that is being considered by SKB, Sweden, for the encapsulation of spent fuel for geological disposal. The principal aims of the study were as follows: a. to determine the mechanisms by which gas generated by anaerobic corrosion will migrate from a canister; b. to identify the possible consequences of gas generation, for example overpressurization of the canisters and effects on water movement; c. to carry out studies to assess the magnitudes of the consequences of gas generation and the way in which they are influenced by the mechanisms and ease of gas migration; d. to determine the likely fate of the gas produced in the repository; for example whether the gas will eventually be dissolved in the groundwater as it moves away from the canister or whether it will collect as free gas in the tunnel or elsewhere; e. to identify the potential benefits of using computer modelling techniques for estimating hydrogen generation rates within disposal canisters during the post-emplacement period

  4. Hanford K basins spent nuclear fuel project update

    International Nuclear Information System (INIS)

    Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building

  5. 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

  6. Corrosion of the copper canister in the repository environment

    International Nuclear Information System (INIS)

    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 there is

  7. STS-45 ATLAS-1 pallets and SSBUV canisters in OV-104's payload bay (PLB)

    Science.gov (United States)

    1992-01-01

    STS-45 payload bay (PLB) configuration onboard Atlantis, Orbiter Vehicle (OV) 104, includes the Shuttle Solar Backscatter Ultraviolet 4 (SSBUV-4) and Atmospheric Laboratory for Applications and Science 1 (ATLAS-1) instruments. The SSBUV get away special (GAS) canisters are mounted on a GAS adapter beam on the starboard PLB sill longeron. THE SSBUV support canister is in the foreground and the SSBUV instrument canister with motorized door assembly (MDA) is next to it. ATLAS-1 equipment includes the igloo (center - decorated with several insignias), the Space Experiments with Particle Accelerators (SEPAC) spheres, and additional instruments mounted on unpressurized spacelab pallets. In the background, are the orbital maneuvering system (OMS) pods and vertical tail highlighted against the cloud-covered surface of the Earth.

  8. 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.

  9. Design, Manufacturing, and Performance estimation of a Disposal Canister for the Ceramic Waste from Pyroprocessing

    International Nuclear Information System (INIS)

    A pyroprocess is currently being developed by KAERI to cope with a highly accumulated spent nuclear fuel in Korea. The pyroprocess produces a certain amount of high-level radioactive waste (HLW), which is solidified by a ceramic binder. The produced ceramic waste will be confined in a secure disposal canister and then placed in a deep geologic formation so as not to contaminate human environment. In this paper, the development of a disposal canister was overviewed by discussing mainly its design premises, constitution, manufacturing methods, corrosion resistance in a deep geologic environment, radiation shielding, and structural stability. The disposal canister should be safe from thermal, chemical, mechanical, and biological invasions for a very long time so as not to release any kind of radionuclides.

  10. Testing of candidate waste-package backfill and canister materials for basalt

    International Nuclear Information System (INIS)

    The Basalt Waste Isolation Project (BWIP) is developing a multiple-barrier waste package to contain high-level nuclear waste as part of an overall system (e.g., waste package, repository sealing system, and host rock) designed to isolate the waste in a repository located in basalt beneath the Hanford Site, Richland, Washington. The three basic components of the waste package are the waste form, the canister, and the backfill. An extensive testing program is under way to determine the chemical, physical, and mechanical properties of potential canister and backfill materials. The data derived from this testing program will be used to recommend those materials that most adequately perform the functions assigned to the canister and backfill

  11. Enhanced Earthquake-Resistance on the High Level Radioactive Waste Canister

    International Nuclear Information System (INIS)

    In this paper, the earthquake-resistance type buffer was developed with the method protecting safely about the earthquake. The main parameter having an effect on the earthquake-resistant performance was analyzed and the earthquake-proof type buffer material was designed. The shear analysis model was developed and the performance of the earthquake-resistance buffer material was evaluated. The dynamic behavior of the radioactive waste disposal canister was analyzed in case the earthquake was generated. In the case, the disposal canister gets the serious damage. In this paper, the earthquake-resistance buffer material was developed in order to prevent this damage. By putting the buffer in which the density is small between the canister and buffer, the earthquake-resistant performance was improved about 80%

  12. Mechanical analysis of cylindrical part of canisters for spent nuclear fuel

    International Nuclear Information System (INIS)

    This report describes mechanical analyses of cylindrical part of the VVER 440-, BWR and EPR-type canisters for spent nuclear fuel. The task was first to evaluate the stresses at maximum design pressure and further by increasing pressure load to determine the limit collapse load and corresponding safety factor. Maximum design pressure 44 MPa is a sum of the hydrostatic pressure 30 MPa caused by 3 km ice layer, 7 MPa caused by ground water pressure at the deepest disposal depth of 700 m and 7 MPa from bentonite swelling pressure. The analysis presented in this report concern the middle area of the canisters, where the cast iron insert is considered to be more critical than in the ends of the canister. For the model a piece from the middle area of the canister was separated by two planes perpendicular to the axis of the canister. This piece was studied first by two-dimensional plane strain model, where the planes are constrained and no elongation of the canister takes place. In the second model one of the planes was constrained and the other plane was allowed to displace in axial direction, which remains as a plane during deformation and to which axial pressure force is directed. This analysis, which corresponds better the real condition in the canister, was performed as threedimensional. The analyses gave however practically equal results due to plastic deformation. Thus the analysis can be done by two-dimensional plane strain model leading to same accuracy with less computation effort. Analyses were performed as large displacement and large strain analyses by the PASULA computing package, which has been developed at VTT for a variety of structural analysis and for heat conduction calculations. A special routine was developed for automatic mesh generation. Before the analysis of the VVER 440-, BWR- and EPR-type canisters the calculation methodology was validated with test results, which were received from pressure tests performed with a short BWR canister in Germany

  13. Canisters for spent-fuel disposal: Design measures against localized corrosion

    International Nuclear Information System (INIS)

    Common to all high-level-waste disposal concepts is the encapsulation of the waste into metal canisters. The purpose of this waste canister is to isolate the radioactive waste from contact with its surroundings for a desired time period. The design service life ranges from hundreds to thousands of years depending on the disposal concept. After the isolation has been breached, other barriers in the disposal system will delay and attenuate the radioactive releases to acceptable levels. In a deep geologic repository, the waste package will be exposed to chemical attack and, depending on the type of repository, to mechanical stresses. Each of these factors will by itself or in combination inevitably lead to loss of confinement some time in the future. In the design of the Swedish waste canister, the corrosion resistance is provided by an outer shell of pure copper while an insert supplies the mechanical strength cast nodular iron. The close fit between the insert and the copper results in very small tensile stresses in the copper over very limited areas once the repository has been saturated. Measurements of stress corrosion crack growth show that annealed copper cannot maintain sufficiently high stress intensity factors for cracks to grow. For annealed copper, the stress intensity factor was limited to 25 MPa·m1/2 because of extensive plastic deformation. For cold-worked copper, no crack growth could be observed for stress intensity factors 1/2. Through the choices of canister material, canister, and repository design, and considering the expected chemical conditions, the risks for localized corrosion can be lowered to an acceptable level, if not eliminated altogether, and the releases from prematurely failed canisters can be kept well within acceptable dose levels

  14. 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.

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

    International Nuclear Information System (INIS)

    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 105 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 105 years

  16. Design basis for the copper/steel canister. Stage three. Final report

    International Nuclear Information System (INIS)

    The development of the copper/iron canister proposed for the containment of high-level waste in the Swedish disposal programme has been studied from the points of view of choice of materials, manufacturing technology and Q A. This report describes the observations on progress which has been made between March 1995 and February 1996 and the results of further literature studies. A first trial canister has been produced by SKB using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. It is considered that such a change will require a significant development programme. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. An improved microstructure may be achieved by extruding at a lower temperature but this remains to be demonstrated. Similar problems exist with plate used for the fabricated tubular but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. However it was necessary to constrain it during welding and it subsequently distorted during machining. There was some evidence of hot tearing close to the weld. The distortion problem may be overcome by a stress relieving anneal but this could cause further grain size problems. 19 refs

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

    International Nuclear Information System (INIS)

    A model has been developed to study the effects of secondary water radiolysis caused by dispersed radionuclides in a bentonite buffer surrounding a copper canister. The secondary radiolysis is the radiolysis caused by radionuclides that have been released from the spent fuel and are present either as solutes in the pore-water, as sorbed species on the surface of other minerals, or as secondary minerals. 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 dissolution of the spent fuel is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. Two cases have been considered with the purpose to illustrate the behaviors of both conservative and non-conservative nuclides. The nuclides that are most relevant are those expected to be the dominant α-emitters in the long-term (e.g. 239Pu, 240Pu, 241Am). in the first case it is assumed that there is no precipitation of secondary minerals of the relevant radionuclides inside the canister. In the second case it is assumed that the radionuclide concentration within the canister is controlled by its respective solubility limit. The radionuclide released to the surrounding buffer is then predicted using a mass balance model. The modelling results show that in both cases, the spent fuel will not be oxidized at a rate significantly faster compared to the case where secondary radiolysis is completely neglected. In the first case, however, a large domain of the near-field can be oxidized due to a much faster depletion of reducing minerals in the buffer, compared to the case where secondary radiolysis is neglected. In the second case, the effects of the secondary water radiolysis will be quite limited. Copyright (2001) Material Research Society

  18. Analyses of atmospheric radon 222 / canisters exposed by Greenpeace in Niger (Arlit / Akokan sector)

    International Nuclear Information System (INIS)

    The companies SOMAIR and COMINAK, subsidiaries of the AREVA group, are mining uranium deposits in northern Niger. In the course of a field mission carried out in November 2009, a Greenpeace International team deposited detectors (canisters of activated charcoal) to measure radon 222, a radioactive gas formed by the decay of the radium 226 present in the uranium ore. This report includes the results of the analysis of the activated charcoal canisters conducted in CRIIRAD's laboratory, and a brief commentary on the interpretation of the results. (authors)

  19. Results of Stainless Steel Canister Corrosion Studies and Environmental Sample Investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R; Enos, David

    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

  20. 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

  1. An evaluation of dual-purpose canisters in the Civilian Radioactive Waste Management System

    International Nuclear Information System (INIS)

    An evaluation was made of the Civilian Radioactive Waste Management System (CRWMS) using dual-purpose canisters (DPCs) and was compared to a system using multi-purpose canisters (MPCs). The DPC would be designed for transportation and storage, whereas the MPC is designed for transportation, storage, and geologic disposal. Implementation of the DPC concept could allow the federal government to proceed with storage and transportation of spent nuclear fuel (SNF) without linkage to geologic disposal, while continuing to independently develop ultimate geologic disposal requirements and designs

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

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. Creep properties of welded joints in copper canisters for nuclear waste containment

    International Nuclear Information System (INIS)

    Copper canisters for nuclear waste containment can be expected to be exposed to temperatures up to 1000C. Since the material is pure copper, creep properties must be taken into account in particular for the welded joints in the canisters. In the paper creep rupture properties of parent metal, weld metal, and simulated heat affected zones are presented for 1100C. About ten times shorter rupture times were found for the weld metal in comparison to the parent metal. Cross weld specimens showed even shorter rupture times

  5. Integrity of copper/steel canisters under crystalline bedrock repository conditions

    International Nuclear Information System (INIS)

    In the Swedish nuclear waste disposal programme, the need to store the spent nuclear fuel safely for very long times has prompted a strategy which includes a long life canister. Technical as well as economical considerations related to design, choice of materials and manufacturing technology have lead to the selection of a reference design to be used for the continued development work. The canisters are cylindrical with a diameter close to 1 meter and a height of about 5 meters. In order to meet the need for an appropriate combination of mechanical strength, toughness, durability and corrosion resistance, the canisters comprise an inner vessel made of steel or cast iron to cope with mechanical stresses and an outer vessel made of almost pure copper to provide corrosion resistance. The Swedish nuclear industry has recently extended its development work to full-scale tests. Such experience is needed not least for the evaluation of the long-term integrity of the canister. This work has been closely followed by the Swedish Nuclear Power Inspectorate (SKI) who have also carried out independent investigations and analyses. It should be emphasized that the findings relate to a canister which is under development and cannot, in general, be expected to be relevant for the fully developed canister. Significant results of the analyses include the identification of conceivable modes of canister failures. Such failures may be related to defects, segregation, limitations in inspectability, long term creep properties, adverse mechanical load situations, etc. It is assessed that the distribution functions of these failures might have their largest uncertainties at the tails extending to comparatively short times. Specific issues related to canister manufacture, scaling and non destructive testing which have been found to warrant further investigation are: defects in the copper ingot which may transfer to the rolled copper plate; the amount of work applied during the rolling or

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

    International Nuclear Information System (INIS)

    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

  7. Numerical analysis of natural convection heat transfer in the shielded canister for the spent fuel

    International Nuclear Information System (INIS)

    PHOENICS-3.2, a three-dimension CFD code is used to research the natural convection heat transfer characters in the horizontal dry shielded canister for the spent fuel assemblies. The computational results are compared with the published experimental and computational results. The results are satisfactory. The parameters of 200 MW Nuclear Heating Reactor are used in the calculations to study the feasibility of the dry shielded canister's application in Nuclear Heating Reactor. Nitrogen and water are chosen as working fluid. In comparison of the heat transfer results of these two kinds of working fluids, nitrogen and water it is found that water is the better choice for Nuclear Heating Reactor

  8. Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

    International Nuclear Information System (INIS)

    The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

  9. Spent Nuclear Fuel project stage and store K basin SNF in canister storage building functions and requirements. Revision 1

    International Nuclear Information System (INIS)

    This document establishes the functions and requirements baseline for the implementation of the Canister Storage Building Subproject. The mission allocated to the Canister Storage Building Subproject is to provide safe, environmentally sound staging and storage of K Basin SNF until a decision on the final disposition is reached and implemented

  10. The simulation and anlaysis on the radioprotection of the TH-PPL CT's lead canister by Monte Carlo method

    International Nuclear Information System (INIS)

    The TH-PPL CT teaching instrument, developed to Tsinghua University, adopts a 137Cs standard radiation source, which is shielded by one lead canister. This paper simulates and analyses the irradiation rate around the lead canister by a method, which combines Monte Carlo and practical measurement. The simulative result validates the correctness of this method. ICRU sphere's sediment energy is simulated, when the ICRU sphere is 50 mm far away from the lead canister. The personal dose will be calculated from the previous step, the results approve that the lead canister's protection is safe and Monte Carlo can be used in radioprotection analysis and optimum design of lead canister to shield radiation source. (authors)

  11. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or

  12. 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.

  13. Experimental assessment of the thermal performance of storage canister/holding fixture configurations for the Los Alamos Nuclear Materials Storage Facility

    International Nuclear Information System (INIS)

    This report presents experimental results on the thermal performance of various nested canister configurations and canister holding fixtures to be used in the Los Alamos Nuclear Materials Storage Facility. The experiment consisted of placing a heated aluminum billet (to represent heat-generating nuclear material) inside curved- and flat-bottom canisters with and without holding plate fixtures and/or extended fin surfaces. Surface temperatures were measured at several locations on the aluminum billet, inner and outer canisters, and the holding plate fixture to assess the effectiveness of the various configurations in removing and distributing the heat from the aluminum billet. Results indicated that the curved-bottom canisters, with or without holding fixtures, were extremely ineffective in extracting heat from the aluminum billet. The larger thermal contact area provided by the flat-bottom canisters compared with the curved-bottom design, greatly enhanced the heat removal process and lowered the temperature of the aluminum billet considerably. The addition of the fixture plates to the flat-bottom canister geometry greatly enhances the heat removal rates and lowers the canister operating temperatures considerably. The addition of the fixture plates to the flat-bottom canister geometry greatly enhances the heat removal rates and lowers the canister operating temperatures considerably. Finally, the addition of extended fin surfaces to the outer flat-bottom canister positioned on a fixture plate, reduced the canister temperatures still further

  14. Development of single tubing-type canister for cryo-storage of bull semen and their effect on sperm motility and viability

    Directory of Open Access Journals (Sweden)

    Mohd Iswadi Ismail

    2014-04-01

    Full Text Available The objective of this study was to evaluate the potential of using single tubing-type canister on sperm quality. Semen was collected from the Bali cattle bull by electroejaculation technique and was cryopreserved in liquid nitrogen using slow freezing cryopreservation method. Two type of canister volume was used in this study; commercial canister (342.25π x 278 mm² and single tubing-type canister (4π x 90 mm². Makler counting chamber and computer assisted sperm analyzer (CASA were used to evaluate the sperm motility and viability of post-thaw sperm. Results showed that the bull sperm motility and viability at the bottom of tubing-type canister was statistically higher and significant as compared to the commercial canister (p<0.05. Significant changes were found in sperm kinetics (VCL, VAP, VSL of tubing-type canister compared to commercial canister. No significant changes in the motility and viability of the bull sperm at the top of tubing-type canister and commercial canister. There were no significant changes in sperm progression (LIN, WOB, PROG in both the canisters. Developed tubing-type canister in this study showed potential as an alternative to be used in bull sperm cryo-storage.

  15. Investigation of exchange bias in 0.1MFe2O4/0.9BiFeO3 (M=Co, Cu, Ni) nanocomposite

    International Nuclear Information System (INIS)

    The 0.1MFe2O4/0.9BiFeO3 (M=Co, Cu, Ni) nanocomposite samples were synthesized by the sol–gel method. Phase composition analysis was carried out, which showed that these bulk samples were composed of a ferrimagnetic MFe2O4 (M=Co, Cu, Ni) and a ferroelectric antiferromagnet (FEAF) BiFeO3 phases, respectively. The magnetic properties of all the samples were investigated by measuring their magnetization as a function of temperature and magnetic field. These results indicated that the magnetic hysteresis loops of 0.1CuFe2O4/0.9BiFeO3 sample sintered in air atmosphere at 550 °C for 3 h exhibited a negative shift and an enhanced coercivity at low temperature ascribed to strong exchange coupling between the BiFeO3 and CuFe2O4 grains. However, there were no magnetic hysteresis loops in both the 0.1CoFe2O4/0.9BiFeO3 sample and the 0.1NiFe2O4/0.9BiFeO3 sample. In view of these results, we tend to think the CuFe2O4/BiFeO3 nanocomposite system may be a useful multifunctional material. - Highlights: ► Exchange bias effect in ferroelectric antiferromagnet (FEAF)/ferromagnet (FM) nanocomposites. ► Exchange bias effect is only observed in the 0.1CuFe2O4/0.9BiFeO3 nanocomposite. ► Lower saturation magnetization is important for producing exchange bias in FEAF/FM system.

  16. Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB) Process Flow Diagram Mass Balance Calculations

    International Nuclear Information System (INIS)

    The purpose of these calculations is to develop the material balances for documentation of the Canister Storage Building (CSB) Process Flow Diagram (PFD) and future reference. The attached mass balances were prepared to support revision two of the PFD for the CSB. The calculations refer to diagram H-2-825869

  17. System design description for the consolidated sludge sampling system for K Basins floor and fuel canisters

    International Nuclear Information System (INIS)

    This System Design Description describes the Consolidated Sludge Sampling System used in the gathering of sludge samples from K Basin floor and fuel canisters. This document provides additional information on the need for the system, the functions and requirements of the systems, the operations of the system, and the general work plan used in its' design and development

  18. 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.

  19. Transport from the canister to the biosphere: Using an integrated near- and far-field model

    International Nuclear Information System (INIS)

    A coupled model concept which may be used for performance assessment of a nuclear repository is presented. The tool is developed by integration of two models, one near field and one far field model. A compartment model, NUCTRAN, is used to calculate the near field release from a damaged canister. The far field transport through fractured rock is simulated by using CHAN3D, based on a three-dimensional stochastic channel network concept. The near field release depends on the local hydraulic properties of the far field. The transport in the far field in turn depends on where the damaged canister(s) is located. The very large heterogeneities in the rock mass makes it necessary to study both the near field release properties and the location of release at the same time. In order to demonstrate the capabilities of the coupled model concept it is applied on a hypothetical repository located at the Hard Rock Laboratory in Aespoe, Sweden. Two main items were studied; the location of a damaged canister in relation to fracture zones and the barrier function of the host rock. In the study of the near field rock as a transport barrier the effect of different tunnel excavation methods which may influence the damage level of the rock around the tunnel was addressed

  20. Gap Analysis to Support Modeling the Long-Term Degradation of Used Nuclear Fuel Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Philip J.; Sunderland, Dion J.; Ross, Steven B.; Montgomery, Robert O.; Hanson, Brady D.; Devanathan, Ram

    2015-04-01

    Welded stainless steel canisters are being used worldwide for dry storage of used nuclear fuel (UNF) assemblies, and the number of canisters in use is steadily increasing. In support of work currently being pursued at Pacific Northwest National Laboratory to understand the atmospheric corrosion behavior of spent fuel dry storage systems, a gap analysis is underway to assess the state of knowledge for modeling of the long-term degradation of a UNF canister. The fundamental aim of this work is to inform research and development (R&D) efforts to establish a sound technical basis to support the extended dry storage of UNF for 100+ years. The analysis is considering all major components of the atmosphere corrosion degradation processes, ranging from contaminant sources and climatic interactions to regional conditions of particle transport and deposition, to microscale effects leading to stress corrosion cracking. The results of this gap analysis will be used to define the R&D pathway to develop an integrated multi-scale atmospheric corrosion modeling capability for UNF in dry storage canisters that can support the safe and reliable performance of these structures for more than 100 years.

  1. Quality Assurance Program Plan for Project W-379: Spent Nuclear Fuels Canister Storage Building Projec

    International Nuclear Information System (INIS)

    This document describes the Quality Assurance Program Plan (QAPP) for the Spent Nuclear Fuels (SNF) Canister Storage Building (CSB) Project. The purpose of this QAPP is to control project activities ensuring achievement of the project mission in a safe, consistent and reliable manner

  2. Analysis of grain boundary corrosion in canister material for radioactive waste using transmission electron microscope

    International Nuclear Information System (INIS)

    Canister for the processed waste is sensitive to corrosion. The grain boundary corrosion is a localized corrosion type which probably takes place on AISI 304 stainless steel canister as a result of pouring the waste glass into it. This research was aimed to study AISI 304 stainless steel as candidate material for high level waste canister. A study of Cr23C6 precipitation at the grain boundary as corrosion initiating agent had been done by observation using Transmission Electron Microscope (TEM). The experiment was carried out by heating the samples at temperature of 700oC for 2 hours followed by water quenching. It was found that the Cr23C6 precipitation occurs and the diameter of the precipitates is 0.2 μm, FCC structure with lattice parameter of 10.585 A. The precipitate was separated one another. It could be said that by the treatment mentioned above, the grain boundary corrosion was insignificant. Therefore the use of AISI 304 stainless steel as canister material candidate of high level waste will be safe from grain boundary corrosion. (author)

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

    International Nuclear Information System (INIS)

    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/

  4. Stress analysis of glass-canister interaction: a study of residual stresses and fracturing

    International Nuclear Information System (INIS)

    Residual stresses and cracking in canisters filled with vitrified nuclear waste are simulated using finite element computer calculations. Cooling rates, internal heat generation, and thermal expansion coefficients significantly affect stress levels. Glass behavior within the softening temperature range is taken to follow the instant freezing concept of Bartenev

  5. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a join ted arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  6. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  7. 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.

  8. The Characteristics of Welding Joint on Stainless Steel as a Candidate of High Level Waste Canister

    International Nuclear Information System (INIS)

    High level waste is the waste generated from reprocessing of the spent fuels. This type of waste is vitrified with borosilicate glass to become waste-glass. This waste glass is contained in a canister made of austenitic stainless steel. The canister material is subjected to be welded during fabrication and utilization. The character of the welding joint that is the function of the electrical current used in the welding process have been studied. The strength of the joint is tested mechanically i.e.: the tensile strength and hardness test. The result shows that the higher the current used in welding process, the better the strength of the joint and as well the tensile strength. The optimum current is 110 A. From the hardness test, it was figured that the length of the HAZ area is 14 mm. The material in HAZ area is the hardest compared to the others, it is due to the appearance of the chrome-carbide. The welding of the canister with such a condition, during fabrication as well as during the utilization of the canister for the container of the high level waste with the PWHT process gives better result. (author)

  9. SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

    International Nuclear Information System (INIS)

    The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report

  10. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    International Nuclear Information System (INIS)

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  11. 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/

  12. Data compliation report: K West Basin fuel storage canister liquid samples

    International Nuclear Information System (INIS)

    Sample analysis data from the 222-S Laboratory are reported. The data are for liquid samples taken from spent fuel storage canisters in the 105 K West Basin during March 1995. An analysis and data report from the Special Analytical Studies group of Westinghouse Hanford Company regarding these samples is also included. Data analysis is not included herein

  13. 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.

  14. Probabilistic analysis and material characterisation of canister insert for spent nuclear fuel. Summary report

    International Nuclear Information System (INIS)

    The KBS-3 canister for geological disposal of spent nuclear fuel in Sweden consists of a ductile cast iron insert and a copper shielding. The canister should inhibit release of radionuclides for at least 100,000 years. The copper protects the canister from corrosion whereas the ductile cast iron insert provides the mechanical strength. In the repository the hydrostatic pressure from the groundwater and the swelling pressure from the surrounding bentonite, which in total results in a maximum pressure of 14 MPa, will load the canisters in compression. During the extreme time scales, ice ages are expected with a maximum ice thickness of 3,000 m resulting in an additional pressure of 30 MPa. The maximum design pressure for the KBS-3 canisters has therefore been set to be 44 MPa. A relatively large number of canisters have been manufactured as part of SKB's development programme. To verify the strength of the canisters at this stage of development SKB initiated a project in cooperation with the European commissions Joint Research Centre (JRC), Institute of Energy in Petten in the Netherlands, together with a number of other partners. Three inserts manufactured by different Swedish foundries were used in the project. A large statistical test programme was developed to determine statistical distributions of various material parameters and defect distributions. These data together with the results from stress and strain finite element analysis were subsequently used in probabilistic analysis to determine the probability for plastic collapse caused by high pressure or fracture by crack growth in regions with tensile stresses. The main conclusions from the probabilistic analysis are: 1. At the design pressure of 44 MPa, the probability of failure is insignificant (∼2x10-9). This is the case even though several conservative assumptions have been made. 2. The stresses in the insert caused by the outer pressure are mainly compressive. The regions with tensile stresses are

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

    International Nuclear Information System (INIS)

    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 2O/Cu(OH)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 smaller discontinuities will die once they reach this maximum size. Death of propagating pits will be compounded by the decrease in oxygen flux to the canister as the repository environment becomes anoxic. Surface discontinuities could impact the SCC behaviour either through their effect on the local environment or via stress concentration or intensification. There is no evidence that surface discontinuities will affect the initiation of SCC by ennoblement of the corrosion potential or the formation of locally aggressive conditions. Stress concentration at pits could lead to crack initiation under some circumstances, but the stress intensity factor for the resultant cracks, or for pre-existing crack-like discontinuities, will be smaller than the

  16. Simulation of residual stresses and deformations in electron beam-welded copper canisters

    International Nuclear Information System (INIS)

    This report presents the modelling of residual stresses and deformations of an EB-welded copper canister. Two different mock-up lengths are modelled with the Abaqus FEA program, and the similarity of those results is studied. Canister mock-ups of 450 mm and 915 mm were chosen for the test cases. The heat treatment results presented in Taskinen 2009 are used as input data for the mechanical model. For the mechanical analysis some simplifications were made to the model. The contact surface between pipe and lid is assumed to be tied and support from the bottom surface is provided with four support points. Results show that, due to the similarity of 450 mm and 915 mm canisters, the short mock-up can be used to predict the stresses and deformation on a full-length canister (5000 mm). The similarity of the temperature fields has already been shown in the previous reports (Taskinen 2009). The main result in the deformation is the shape of the canister in the residual state. The top of the canister tries to shrink, resulting in the lid buckling inwards. The deformation of the lid of the canister is about 2.2 mm at the centre of the lid. The main results in the stresses are the stress level on the surface, the deviation of stresses over the circle and the stresses near the welding. On the surface there are areas where the circumferential stress is at tension. However, radial and axial stresses are usually in compression on the surface. The deviation of the stress level over the circle is quite small, except in the overlap area and near it. The residual stresses from 0 deg C to 45 deg C change remarkably, but over the rest of the area the stresses are more constant. Near the welding the stresses on the top surface are in compression, but in the centre of the welding the stresses are in tension. In the modelling, the possibility of calculating a mechanical model with the contact surface between pipe and lid, so that they could be separated during the welding, was also tested

  17. Development of fabrication technology for copper canisters with cast inserts. Status report in August 2001

    International Nuclear Information System (INIS)

    This report contains an account of the results of trial fabrication of copper canisters with cast inserts carried out during the period 1998 - 2001. The work of testing of fabrication methods is being focused on a copper thickness of 50 mm. Occasional canisters with 30 mm copper thickness are being fabricated for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. For the fabrication of copper tubes, SKB has concentrated its efforts on seamless tubes made by extrusion and pierce and draw processing. Five tubes have been extruded and two have been pierced and drawn during the period. Materials testing has shown that the resultant structure and mechanical properties of these tubes are good. Despite certain problems with dimensional accuracy, it can be concluded that both of these methods can be developed for use in the serial production of SKB' copper tubes. No new trial fabrication with roll forming of copper plate and longitudinal welding has been done. This method is nevertheless regarded as a potential alternative. Copper lids and bottoms are made by forging of continuous-cast bars. The forged blanks are machined to the desired dimensions. Due to the Canister Laboratory's need for lids to develop the technique for sealing welding, a relatively large number of forged blanks have been fabricated. It is noted in the report that the grain size obtained in lids and bottoms is much coarser than in fabricated copper tubes. Development work has been commenced for the purpose of optimizing the forging process. Nine cast inserts have been cast during the three-year period. The results of completed material testing of test pieces taken at different places along the length of the inserts have in several cases shown an unacceptable range of variation in strength properties and structure. In the continued work, insert fabrication will be developed in terms of both casting technique and iron composition. Development work on

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

    International Nuclear Information System (INIS)

    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

  20. 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

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

    International Nuclear Information System (INIS)

    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 obtained in

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

    International Nuclear Information System (INIS)

    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 obtained in

  3. Simulation of long-term behavior in HLW near-field by centrifugal model test. Part 4. Model test of coupled THM processes in isotropic stress conditions using heatable overpack

    International Nuclear Information System (INIS)

    We demonstrated the equivalent long-term behavior in the near-field of a geological repository for high level radioactive waste disposal, using the centrifugal near-field model test under the coupled thermo-hydraulic-mechanical condition. The model consisted of a sedimentary bedrock, buffer, and heating type model overpack, and was enclosed within a pressure vessel. Tests were conducted with a centrifugal force field of 30 G under isotropic stress-constrain conditions with confining pressures and injection of pore water. The temperature condition of the overpack was constantly 95°C. As the result, the values showed similar behaviors to that of the normal temperature tests partially. However, the different behaviors were measured such as the displacement of overpack change from the settlement to the heave, the extreme drop in the soil pressure of the buffer and the strain of side wall of bedrock change from the tension to the compression after injecting pore water of hundreds hours. In addition, the flow rate of the injection pore water suddenly changed after hundreds of hours. Furthermore, the density of the buffer was lower than that of the normal temperature tests by X-ray CT imaging in the post-tests. We infer that the high temperature overpack influenced the stiffness and the pore water distribution of the buffer, and the density and the soil pressure of the buffer decreased. As a result of the change of stiffness in the disposal hole (buffer), the tendency to the strain of the surrounding bedrock and the displacement of the overpack changed. (author)

  4. Nonlinear dynamic impact analysis for installing a dry storage canister into a vertical concrete cask

    International Nuclear Information System (INIS)

    In this paper, a series of dynamic impact analysis for installing a dry storage canister into a vertical concrete cask (VCC) is performed. The dry storage system considered herein is called HCDSS-69, recently developed by INER and being capable of accommodating 69 bundles of BWR spent nuclear fuels. The impact accident is stemming from a conservative consideration of accidental movement when the canister is being hoisted into a VCC. According to NUREG-0554, the accidental movement is conservatively simulated by 80 mm- and 160 mm-height free-drop motions and then with straight and 2°-oblique impact to a pedestal in VCC. A symmetric fully 3-D finite element model is built and analyzed using the explicit finite element code, LS-DYNA. Geometrical, contact, and material nonlinearities are all taken into account. The analysis result concludes that the permanent deformations of the canister are not severe to affect fuel retrieve after the impact accident and the maximum stress intensity in the canister shell can meet the ASME code appendix F F-1340, preventing the leakage of radioactive materials. The study also found that with properly reducing the wall thickness of the pedestal cylinder, the maximum acceleration and permanent deformation of the canister can be much alleviated, even though the drop height is increased to the double of the required brake distance specified in NUREG-0554. The damages of the pedestal in each analysis are moderate so that the heat transfer condition after the impact accident can be bounded by the off-normal event for half-blockage of air inlets

  5. 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.

  6. 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

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

    International Nuclear Information System (INIS)

    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

  8. Thermal analysis of dry concrete canister storage system for CANDU spent fuel

    International Nuclear Information System (INIS)

    This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

  9. Spent Nuclear Fuel (SNF) Project Product Specification

    International Nuclear Information System (INIS)

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  10. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  11. Development of measurement technology of chlorine attached on canister using laser. Application of LIBS using collinear geometry

    International Nuclear Information System (INIS)

    A concrete cask is adopted for interim storage of spent fuel. The facility has a natural ventilating system to cool down a stainless steel canister inside the concrete cask. When sea salt particles enter into the ventilating system and attach to the canister, the canister has a possibility to suffer SCC(Stress Corrosion Cracking) induced by chlorine. Therefore, measurement of concentration of chlorine on the canister is requested to check the occurrence of SCC. Laser-induced breakdown spectroscopy (LIBS) is suitable for on-site measurement of concentration of chlorine attached on the canister because noncontact measurement for a canister with high temperature is possible. Experiments were performed using stainless steel plates (SUS304L, SUS316L) sprayed with synthetic seawater. Nd:YAG laser beam was focused onto the SUS304L and SUS316L sample and the emission of the ablated plasma was detected by a spectrometer and an intensified CCD camera. The chlorine spectra were measured for the samples with chlorine concentration from 0.0 g/m2 to 4.0 g/m2 by using single or double pulse measurement. The double pulse measurement was designed by collinear geometry. The intensity of the chlorine fluorescence normalized by oxygen fluorescence increased monotonously versus chlorine concentration from 0.0 to 0.4 g/m2 in double pulse measurements. These results show the possibility of the quantitative measurement of chlorine content on the canister by LIBS. (author)

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

    International Nuclear Information System (INIS)

    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

  13. 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

  14. 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.

  15. Comments on 'SKB FUD-program 95' focused on canister integrity and corrosion

    International Nuclear Information System (INIS)

    The work presented in this report is a result of reading the SKB program for R,D and D on safe storage of radioactive wastes. Our work, which is focused on the waste canisters, was commissioned by the Swedish Nuclear Power Inspectorate. We find the program very difficult to follow owing to the lack of detail in chapter seven. In our opinion this will make the work difficult to monitor by SKI or SKB. We also feel that the interpretation of information already available is overoptimistic. As a consequence the difficulties ahead are understated and the programme is converging too quickly. We believe that it should be possible to develop a satisfactory canister for disposal of high level nuclear waste according to the general method proposed by SKB and with the proposed capacity within the timescale of the overall programme. We do not believe, however, that all the difficulties have been recognised. As a consequence of this the results to date are interpreted optimistically. We believe that progress should be subjected to more professional review within SKB and that a higher level of metallurgical support is required. We disagree that suitable full size canisters have been created and that production technology is available for both canisters at full size. We also disagree that the long-time durability is ascertained. I.a. it is easy to find corrosion mechanisms for the canister system that have to be demonstrated not to be harmful. We feel there are many areas which need further evaluation, i.a. effects of non uniform loading and creep, effects of departure from circularity, welding, quality control, effects of radiolysis, corrosion properties, etc. We also feel that insufficient emphasis has been placed on the further development on high power electron beam welding, machining, casting of the insert, testing and overall handling. We consider that more information should be provided on the detail and timing of the development plan for the trial fabrication programme of

  16. Sampling and Analysis Plan for canister liquid and gas sampling at 105-KW fuel storage basin

    International Nuclear Information System (INIS)

    This Sampling and Analysis Plan (SAP) details the sampling and analyses to be performed on fuel canisters transferred to the Weasel Pit of the 105-KW fuel storage basin. The radionuclide content of the liquid and gas in the canisters must be evaluated to support the shipment of fuel elements to the 300 Area in support of the fuel characterization studies (Abrefah, et al. 1994, Trimble 1995). The following sections provide background information and a description of the facility under investigation, discuss the existing site conditions, present the constituents of concern, outline the purpose and scope of the investigation, outline the data quality objectives (DQO), provide analytical detection limit, precision, and accuracy requirements, and address other quality assurance (QA) issues

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

    International Nuclear Information System (INIS)

    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

  18. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

  19. A crane is lowered over the payload canister with the SRTM inside

    Science.gov (United States)

    1999-01-01

    A crane is lowered over the payload canister with the Shuttle Radar Topography Mission (SRTM) inside in Orbiter Processing Facility (OPF) bay 2. The primary payload on STS-99, the SRTM will soon be lifted out of the canister and installed into the payload bay of the orbiter Endeavour. The SRTM consists of a specially modified radar system that will gather data for the most accurate and complete topographic map of the Earth's surface that has ever been assembled. SRTM will make use of radar interferometry, wherein two radar images are taken from slightly different locations. Differences between these images allow for the calculation of surface elevation. The SRTM hardware includes one radar antenna in the Shuttle payload bay and a second radar antenna attached to the end of a mast extended 60 meters (195 feet) from the shuttle. STS-99 is scheduled to launch Sept. 16 at 8:47 a.m. from Launch Pad 39A.

  20. Moisture probe using neutron moderation for PuO2 canister inspection

    International Nuclear Information System (INIS)

    At several U.S. Department of Energy sites, where the production of nuclear materials was once active, powdered PuO2 contained in small metal canisters is sealed in larger containers for long-term storage. To prevent corrosion and the generation of significant amounts of hydrogen gas within the small canisters, the moisture content of the PuO2 powder must be 2 powder in situ, we proposed the development of a system that is based on the moderation of neutrons. We discuss the results of calculations and measurements performed in a project supported by Los Alamos National Laboratory (LANL) to examine the capabilities and sensitivity of this inspection technique

  1. 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.

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

    International Nuclear Information System (INIS)

    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

  3. Shielding analysis of depleted uranium silicate filler concept for spent fuel canister designs

    International Nuclear Information System (INIS)

    A Depleted Uranium Silicate Container Backfill System (DUSCOBS) has been proposed at Oak Ridge National Laboratory. This concept suggests the use of small, depleted-uranium silicate glass beads as a backfill material inside storage, transportation, and repository waste packages containing spent nuclear fuel. Use of this backfdl material would substantially reduce external dose rates from a waste canister, allowing a reduction of the amount of external shielding required. This paper summarizes the results of scoping studies to estimate the dose reduction from the use of DUSCOBS in a conceptual canister design, and to determine what design modifications are required to offset the increased mass of the system, while simultaneously maintaining sufficient shielding to meet external dose rate limits

  4. 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.

  5. Magnetic properties of BaTiO3 and BaTi1−xMxO3 (M=Co, Fe) nanocrystals by hydrothermal method

    International Nuclear Information System (INIS)

    BaTiO3 and BaTi1−xMxO3 (M=Co, Fe) nanocrystals were prepared by hydrothermal method. X-ray diffraction analysis indicated that all of the samples were of single-phase with tetragonal perovskite structure. The BaTiO3 prepared exhibited weak ferromagnetism rather than diamagnetism, probably due to the oxygen vacancies at the surface. Paramagnetism was observed for all BaTi1−xCoxO3 samples with 0.05≤x≤0.25. The Curie-Weiss fit revealed the paramagnetic moment per Co ion were 4.09 μB, 4.12 μB, and 4.36 μB for x=0.15, 0.20, and 0.25 respectively. Room temperature hysteresis loops of the Fe-doped BaTiO3 samples were observed at the doping level x between 0.2 and 0.5. The saturation magnetization firstly increased with increasing Fe content, but gradually decreased. The divergence was observed in the temperature dependence of the field cooling (FC) and zero-FC (ZFC) magnetization curves, indicating a spin-glass behavior arising from micromagnetic state, i.e. the mixing of ferromagnetic, and antiferromagnetic phases. The observed ferromagnetism may originate from the coupling between the secondary-nearest Fe ions and the antiferromagnetism may be due to the coupling between the nearest Fe ions. The ferromagnetic coupling competes with the antiferromagnetic coupling. Therefore, the ferromagnetic properties are predominant when the Fe doping level are at a certain range. - Highlights: • We prepare BaTiO3 and BaTi1−xMxO3 (M=Co, Fe) by the hydrothermal method. • BaTiO3 nanocrystals by hydrothermal method exhibit weak ferromagnetism. • BaTi1−xCoxO3 (0.05≤x≤0.25) nanocrystals exhibit paramagnetism. • BaTi1−xFexO3 (0.10≤x≤0.50) nanocrystals exhibit ferromagnetism

  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. Feasibility of using a high-level waste canister as an engineered barrier in disposal

    International Nuclear Information System (INIS)

    The objective of this report is to evaluate the feasibility of designing a process canister that could also serve as a barrier canister. To do this a general set of performance criteria is assumed and several metal alloys having a high probability of demonstrating high corrosion resistance under repository conditions are evaluated in a qualitative design assessment. This assessment encompasses canister manufacture, the glass-filling process, interim storage, transportation, and to a limited extent, disposal in a repository. A series of scoping tests were carried out on two titanium alloys and Inconel 625 to determine if the high temperature inherent in the glass-fill processing would seriously affect either the strength or corrosion resistance of these metals. This is a process-related concern unique to the barrier canister concept. The material properties were affected by the heat treatments which simulated both the joule-heated glass melter process (titanium alloys and Inconel 625) and the in-can melter (ICM) process (Inconel 625). However, changes in the material properties were generally within 20% of the original specimens. Accelerated corrosion testing of the heat treated coupons in a highly oxygenated brine showed basic corrosion resistance of titanium grade 12 and Inconel 625 to compare favorably with that of the untreated coupons. The titanium grade 2 coupons experienced severe corrosion pitting. These corrosion tests were of a scoping nature and suitable primarily for the detection of gross sensitivity to the heat treatment inherent in the glass-fill process. They are only suggstive of repository performance since the tests do not adequately model the wide range of repository conditions that could conceivably occur

  8. Analytical Results of DWPF Glass Sample Taken During Filling of Canister S01913

    International Nuclear Information System (INIS)

    The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 2 (SB2) in December 2001 as part of Sludge Receipt and Adjustment Tank (SRAT) Batch 208. Sludge Batch 2 consists of the contents of Tank 40 and Tank 8 in approximately equal proportions. The sludge slurry is received into the DWPF Chemical Processing Cell and is processed through the SRAT and Slurry Mix Evaporator Tank. The treated sludge slurry is then transferred to the Melter Feed Tank and fed to the melter. During the processing of each sludge batch, the DWPF is required to take at least one glass sample. This glass sample is taken to meet the objectives of the Glass Product Control Program1 and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository.The DWPF requested analysis of a radioactive glass sample obtained from the melter pour stream during the processing of Macrobatch 3 (MB3) (Sludge Batch 2)2 with Frit 320. A glass sample was obtained while pouring Canister S01913 and was sent to the Savannah River National Laboratory Shielded Cells for characterization. Canister S01913 was the 267th canister poured during vitrification of Sludge Batch 2 (364 canisters of glass were prepared from SB2). The glass sample arrived from DWPF in primary container PC0034. The primary container contained pieces of glass. The glass had been extracted from the sample holder in the DWPF. This report contains the visual observations of the as-received glass sample, results for the density, the chemical composition, the Product Consistency Test and the calculated and measured radionuclide results needed for the Production

  9. Monitored Retrievable Storage/Multi-Purpose Canister analysis: Simulation and economics of automation

    International Nuclear Information System (INIS)

    Robotic automation is examined as a possible alternative to manual spent nuclear fuel, transport cask and Multi-Purpose canister (MPC) handling at a Monitored Retrievable Storage (MRS) facility. Automation of key operational aspects for the MRS/MPC system are analyzed to determine equipment requirements, through-put times and equipment costs is described. The economic and radiation dose impacts resulting from this automation are compared to manual handling methods

  10. 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

  11. Comparative evaluations of the thermomechanical responses for three high level waste canister emplacement alternatives

    International Nuclear Information System (INIS)

    The structural responses of three room and canister configurations proposed for the underground storage of high level nuclear wastes have been compared. Coupled secondary creep and heat transfer computations indicate that the future retrieval of waste is most readily assured with a design that combines a low extraction ratio (large pillars) with waste emplacement into the floors of each storage room. Thermoelastic computations show minimal room closure in comparison to room closure due to creep deformations

  12. A charcoal canister survey of radon emanation at the rehabilitated uranium mine site at Nabarlek

    International Nuclear Information System (INIS)

    This paper describes a recent survey of radon emanation measurements from the rehabilitated Nabarlek mine site. It was mined out in 1979, decommissioned in 1995 and provided a good test bed for assessment of rehabilitation in terms of radon flux attenuation. Measurements have been made with charcoal canisters. Studies to measure the radon-220 flux by observing Tl-208 progeny of thoron the effectiveness of trial covers and meteorological considerations will be reported

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

    International Nuclear Information System (INIS)

    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

  14. Automated waste canister docking and emplacement using a sensor-based intelligent controller

    International Nuclear Information System (INIS)

    A sensor-based intelligent control system is described that utilizes a multiple degree-of-freedom robotic system for the automated remote manipulation and precision docking of large payloads such as waste canisters. Computer vision and ultrasonic proximity sensing are used to control the automated precision docking of a large object with a passive target cavity. Real-time sensor processing and model-based analysis are used to control payload position to a precision of ± 0.5 millimeter

  15. 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.

  16. 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.

  17. ALPHN: A computer program for calculating (α, n) neutron production in canisters of high-level waste

    International Nuclear Information System (INIS)

    The rate of neutron production from (α, 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 (α, 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 (α, n) neutron production of each actinide in neutrons per second and the total for the canister. The (α, 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

  18. Challenge to Overcome the Concern of SCC in Canister During Long-Term Storage of Spent Fuel

    International Nuclear Information System (INIS)

    In order to put the concrete cask in practical use in Japan (an island country), stress corrosion cracking (SCC) of canister must be coped with. It is required to take measures for one or two of the three factors, i.e. welding residual stress, material, and environment, to cope with the SCC that may result in loss of the containment function of the canister. Prevention of loss of containment due to SCC of a canister was evaluated either by a method of comparing the amount of salt on the canister surface during storage with the minimum amount of salt to initiate rust and SCC or by a method of comparing the wetting time of the canister surface under salty-air field environment with the lifetime of the SCC fracture of the canister material. Although the use of highly corrosion-resistance stainless steel is one solution, it brings about a cost rise of the concrete cask storage. In order to suppress the cost rise, it should be evaluated whether the measure against SCC of the normal stainless steel is possible by reducing welding residual stress. In addition, technology should be developed to reduce salt particles in the air flowing into the storage facility and concrete cask. (author)

  19. Analysis of burns caused by pre-filled gas canisters used for lamps or portable camping stoves.

    Science.gov (United States)

    Desouches, C; Salazard, B; Romain, F; Karra, C; Lavie, A; Volpe, C Della; Manelli, J C; Magalon, G

    2006-12-01

    The use of pre-filled valveless gas canisters for lamps or camping stoves has caused a number of serious burn incidents. We performed a retrospective analysis of all of the patients who were victims of such incidents admitted to the Marseille Burn Centre between January 1990 and March 2004. There were a total of 21 patients burned in such conditions. Adult males made up the majority of the victims of this sort. Lesions were often extensive (60% of the patients were burned over more than 10% of their body surface) and systematically deep. In order of frequency, burn locations were: the lower limbs, the upper limbs, the hands and the face. The incidents principally occurred during replacement of the canister near an open flame. The marketing of a canister with a valve in order to avoid gas leaks did not cause the old canisters to be taken off the market. On the contrary, European Safety Standard EN417, updated in October 2003, validated the use of these valveless canisters. The severity of the lesions caused and the existence of safe equivalent products requires the passage of a law that forbids valveless canisters. PMID:16982156

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

    International Nuclear Information System (INIS)

    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.