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Sample records for canister storage building

  1. Canister storage building natural phenomena design loads

    International Nuclear Information System (INIS)

    Tallman, A.M.

    1996-02-01

    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

  2. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Canister storage building trade study. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

    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.

  4. Canister storage building trade study. Final report

    International Nuclear Information System (INIS)

    Swenson, C.E.

    1995-05-01

    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

  5. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

    POWERS, T.B.

    1999-01-01

    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 storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    KOPELIC, S.D.

    1999-01-01

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

  7. 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.

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

    International Nuclear Information System (INIS)

    CROWE, R.D.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

    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

  10. 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.

  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)

    Swenson, C.E.

    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

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

    International Nuclear Information System (INIS)

    Dana, W.P.

    1995-12-01

    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

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

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

  15. Mechanical design of the storage tubes in the HWVP canister storage building

    International Nuclear Information System (INIS)

    Divona, C.J.; Fages, R.; Janicek, G.P.; Mullally, J.A.

    1993-01-01

    Canisters of high-level waste from the Hanford Waste Vitrification Plant (HWVP) will be stored in an adjacent facility, the Canister Storage Building (CSB). The canisters are stored vertically in an array of tubes within the shielded vault area of the CSB. This paper describes the mechanical design of the storage tubes, the shield floor plugs that confine the waste within the tubes and the impact absorber system used to assure that the canisters are not breached in the event of an accidental drop. Installation and testing of the components is also discussed

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

    Energy Technology Data Exchange (ETDEWEB)

    TU, K.C.

    1999-10-08

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

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

    International Nuclear Information System (INIS)

    TU, K.C.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1996-01-01

    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

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

    International Nuclear Information System (INIS)

    Womack, J.C.

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

    Duncan, D.W.

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    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

  2. 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)

    HEARD, F.J.

    1999-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    PICKETT, W.W.

    2000-09-22

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

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

    International Nuclear Information System (INIS)

    PICKETT, W.W.

    2000-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2001-05-15

    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. Revision 1 documented 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 in section 3

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

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2001-01-01

    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. Revision 1 documented 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 in section 3.1.5 and will be

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1997-01-01

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

  11. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2003-02-12

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed.

  12. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2003-01-01

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

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

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

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

    GARRISON, R.C.

    2000-01-01

    This document implements the procedure for providing configuration control for the monitoring and control systems associated with the operation of the Canister Storage Building (CSB). It identifies and defines the configuration items in the monitoring and control systems, provides configuration control of these items throughout the system life cycle, provides configuration status accounting, physical protection and control, and verifies the completeness and correctness of the items. It is written to comply with HNF-SD-SNF-CM-001, Spent Nuclear Fuel Configuration Management Plan (Forehand 1998), HNF-PRO-309, Computer Software Quality Assurance Requirements, HNF-PRO-2778, IRM Application Software System Life Cycle Standards, and applicable sections of administrative procedure AP-CM-6-037-00, SNF Project Process Automation Software and Equipment Configuration Management

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

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1997-04-28

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

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1997-01-01

    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

  18. Radioactive air emissions notice of construction for Canister Storage Building (revised sealing configuration for spent nuclear fuel) - Project W-379

    International Nuclear Information System (INIS)

    Kamberg, L.D.

    1998-01-01

    The purpose of this Notice of Construction (NOC) is to provide a rewritten NOC for obtaining regulatory approval for changes to the previous Canister Storage Building (CSB) NOCs (WDOH, 1996 and EPA, 1996) as were approved by the Washington State Department of Health (WDOH, 1996a) and US Environmental Protection Agency (EPA, 1996a). These changes are because of a revised sealing configuration of the multi-canister overpacks (MCOS) that are used to store the SNF. A flow schematic of the SNF Project is provided in Figure 1-1. A separate notification of startup will be provided apart from this NOC

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

    Energy Technology Data Exchange (ETDEWEB)

    GARVIN, L.J.

    1999-09-20

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

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

    International Nuclear Information System (INIS)

    GARVIN, L.J.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    GARVIN, L.J.

    1999-01-01

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

  2. Waste canister for storage of nuclear wastes

    Science.gov (United States)

    Duffy, James B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall.

  3. Waste canister for storage of nuclear wastes

    International Nuclear Information System (INIS)

    Duffy, J.B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall. 4 claims, 4 figures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-29

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

  5. Design of double containment canister cask storage system

    International Nuclear Information System (INIS)

    Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

    2004-01-01

    Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

  6. Evaluation of canister weld flaw depth for concrete storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Tae Chul; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Jung, Sung Hun; Lee, Young Oh; Jung, In Su [Korea Nuclear Engineering and Service Corp, Daejeon (Korea, Republic of)

    2017-03-15

    Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radioactive materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B and PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

  7. Criticality safety for TMI-2 canister storage at INEL

    International Nuclear Information System (INIS)

    Jones, R.R.; Briggs, J.B.; Ayers, A.L. Jr.

    1986-01-01

    Canisters containing Three Mile Island Unit 2 (TMI-2) core debris will be researched, stored, and prepared for final disposition at the Idaho National Engineering Laboratory (INEL). The canisters will be placed into storage modules and assembled into a storage rack, which will be located in the Test Area North (TAN) storage pool. Criticality safety calculations were made (a) to ensure that the storage rack is safe for both normal and accident conditions and (b) to determine the effects of degradation of construction materials (Boraflex and polyethylene) on criticality safety

  8. High-level radioactive waste glass and storage canister design

    International Nuclear Information System (INIS)

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

    1979-01-01

    Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

  9. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    International Nuclear Information System (INIS)

    Bryan, Charles R.; Enos, David G.

    2015-01-01

    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.

  10. 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.

  11. Shielded Canister Transporter

    International Nuclear Information System (INIS)

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

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

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1978-12-01

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

  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. Numerical analysis of a natural convection cooling system for radioactive canisters storage

    Energy Technology Data Exchange (ETDEWEB)

    Tsal, R.J.; Anwar, S.; Mercada, M.G. [Fluor Daniel Inc., Irvine, CA (United States)

    1995-02-01

    This paper describes the use of numerical analysis for studying natural convection cooling systems for long term storage of heat producing radioactive materials, including special nuclear materials and nuclear waste. The paper explains the major design philosophy, and shares the experiences of numerical modeling. The strategy of storing radioactive material is to immobilize nuclear high-level waste by a vitrification process, convertion it into borosilicate glass, and cast the glass into stainless steel canisters. These canisters are seal welded, decontaminated, inspected, and temporarily stored in an underground vault until they can be sent to a geologic repository for permanent storage. These canisters generate heat by nuclear decay of radioactive isotopes. The function of the storage facility ventilation system is to ensure that the glass centerline temperature does not exceed the glass transition temperature during storage and the vault concrete temperatures remain within the specified limits. A natural convection cooling system was proposed to meet these functions. The effectiveness of a natural convection cooling system is dependent on two major factors that affect air movement through the vault for cooling the canisters: (1) thermal buoyancy forces inside the vault which create a stack effect, and (2) external wind forces, that may assist or oppose airflow through the vault. Several numerical computer models were developed to analyze the thermal and hydraulic regimes in the storage vault. The Site Model is used to simulate the airflow around the building and to analyze different air inlet/outlet devices. The Airflow Model simulates the natural convection, thermal regime, and hydraulic resistance in the vault. The Vault Model, internal vault temperature stratification; and, finally, the Hot Area Model is used for modeling concrete temperatures within the vault.

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

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-08

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

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

    International Nuclear Information System (INIS)

    Heard, F.J.

    1999-01-01

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

  17. Life Prediction of Spent Fuel Storage Canister Material

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, Ronald

    2018-04-16

    The original purpose of this project was to develop a probabilistic model for SCC-induced failure of spent fuel storage canisters, exposed to a salt-air environment in the temperature range 30-70°C for periods up to and exceeding 100 years. The nature of this degradation process, which involves multiple degradation mechanisms, combined with variable and uncertain environmental conditions dictates a probabilistic approach to life prediction. A final report for the original portion of the project was submitted earlier. However, residual stress measurements for as-welded and repair welds could not be performed within the original time of the project. As a result of this, a no-cost extension was granted in order to complete these tests. In this report, we report on the results of residual stress measurements.

  18. Structural assessment of a space station solar dynamic heat receiver thermal energy storage canister

    Science.gov (United States)

    Thompson, R. L.; Kerslake, T. W.; Tong, M. T.

    1988-01-01

    The structural performance of a space station thermal energy storage (TES) canister subject to orbital solar flux variation and engine cold start up operating conditions was assessed. The impact of working fluid temperature and salt-void distribution on the canister structure are assessed. Both analytical and experimental studies were conducted to determine the temperature distribution of the canister. Subsequent finite element structural analyses of the canister were performed using both analytically and experimentally obtained temperatures. The Arrhenius creep law was incorporated into the procedure, using secondary creep data for the canister material, Haynes 188 alloy. The predicted cyclic creep strain accumulations at the hot spot were used to assess the structural performance of the canister. In addition, the structural performance of the canister based on the analytically determined temperature was compared with that based on the experimentally measured temperature data.

  19. Canister Storage Building (CSB) Technical Safety Requirements

    International Nuclear Information System (INIS)

    KRAHN, D.E.

    2000-01-01

    The purpose of this section is to explain the meaning of logical connectors with specific examples. Logical connectors are used in Technical Safety Requirements (TSRs) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TSRs are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

    International Nuclear Information System (INIS)

    Allen, G.K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

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

    International Nuclear Information System (INIS)

    Trimble, D.J.

    1995-01-01

    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

  5. 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.

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

    International Nuclear Information System (INIS)

    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

  7. Multi-purpose canisters as an alternative for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Hollaway, W.R.; Rozier, R.; Nitti, D.A.; Williams, J.R.

    1993-01-01

    A study was conducted to assess the feasibility of using multi-purpose canisters to handle spent nuclear fuel throughout the Civilian Radioactive Waste Management System. Multi-purpose canisters would be sealed, metallic containers maintaining multiple spent fuel assemblies in a dry, inert environment and overpacked separately and uniquely for the various system elements of storage, transportation, and disposal. Using five implementation scenarios, the multi-purpose canister was evaluated with regard to several measures of effectiveness, including number of handlings, radiation exposure, cost, schedule and licensing considerations, and public perception. Advantages and disadvantages of the multi-purpose canister were identified relative to the current reference system within each scenario, and the scenarios were compared to determine the most effective method of implementation

  8. Conceptual design study of a concrete canister spent-fuel storage facility

    International Nuclear Information System (INIS)

    Lidfors, E.D.; Tabe, T.; Johnson, H.M.

    1979-01-01

    This report presents a conceptual design study for the interim storage of CANDU spent fuel in concrete canisters. The canisters will be concrete flasks, which contain fuel prepackaged in double steel containment, and will be cooled by natural air convection. This is one of the methods proposed as a potential alternative to water pool storage. A preliminary study of this concept was done by CAFS (Committee Assessing Fuel Storage), and WNRE (Whiteshell Nuclear Research Establishment) is currently conducting a development and demonstration program. This study of a central facility for the storage of all Canadian spent fuel arisings to the year 2000 was completed in 1975. A brief description of the facilities required and the operations involved, a summary of costs, a survey of the monitoring requirements and a prediction of the personnel exposures associated with this method of storing spent fuel are reported here. The estimated total cost of interim storage in cylindrical canisters at a central site is $6.02/kg U (1975 dollars). Approximately half of this cost is incurred in the shipment of fuel from the reactors to the storage facility. (author)

  9. Dry storage developments in France build on CASCAD experience

    International Nuclear Information System (INIS)

    Bonnet, C.; Giraud, C.

    1992-01-01

    The CASCAD dry store, located at CEA's research centre at Cadarache, stores spent fuel from the EL4 heavy water reactor and the Osiris research reactor. The design was based on the following criteria: Storage period. Interim storage is provided for 50 years. Containment. The fuel is contained by a multiple barrier system consisting of: the fuel canister (primary barrier); the sealed stainless steel storage well; and the storage building which includes a ventilation system to provide dynamic containment during handling operations. The fuel is loaded into canisters at the reactor site to avoid contamination in the storage building. The integrity of the primary barrier is periodically monitored by sampling of air from the storage well. Cooling. The storage wells are cooled by a natural convection system that maintains the temperature of the fuel below its stated limit and the temperature of the concrete below 80 o C. Criticality. Criticality incidents are prevented by static design measures such as maintaining a minimum pitch between storage wells and providing sufficient storage well diameter. Radiation protection. Radiation shielding limits the maximum equivalent dose rate for operating personnel to less than 25μSv/h at the handling cell floor and the wall adjoining the control room, and to less than 7.5μSv/h at the outside walls of the storage building. Cannister design. The canister must resist corrosion caused by condensation as well as pressure due to radiolytic gases. The canister must also withstand a drop of up to 10m without losing its integrity. The design has now been adapted to accommodate light reactor fuels and is known as CASCAD+. (Author)

  10. Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

  11. 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)

    Bernardin, J.D.; Naffziger, D.C.; Gregory, W.S.

    1997-11-01

    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

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

    International Nuclear Information System (INIS)

    Ryu, Yong Ho

    1992-02-01

    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

  13. FY17 Status Report: Research on Stress Corrosion Cracking of SNF Interim Storage Canisters.

    Energy Technology Data Exchange (ETDEWEB)

    Schindelholz, Eric John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alexander, Christopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    This progress report describes work done in FY17 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. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.

  14. 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

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

    International Nuclear Information System (INIS)

    Andrews, R.E.

    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

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

    International Nuclear Information System (INIS)

    Bennett, P.C.; Stringer, J.B.

    1994-01-01

    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

  17. Alternative design concept for the second Glass Waste Storage Building

    International Nuclear Information System (INIS)

    Rainisch, R.

    1992-10-01

    This document presents an alternative design concept for storing canisters filled with vitrified waste produced at the Defense Waste Processing Facility (DWPF). The existing Glass Waste Storage Building (GWSB1) has the capacity to store 2,262 canisters and is projected to be completely filled by the year 2000. Current plans for glass waste storage are based on constructing a second Glass Waste Storage Building (GWSB2) once the existing Glass Waste Storage Building (GWSB1) is filled to capacity. The GWSB2 project (Project S-2045) is to provide additional storage capacity for 2,262 canisters. This project was initiated with the issue of a basic data report on March 6, 1989. In response to the basic data report Bechtel National, Inc. (BNI) prepared a draft conceptual design report (CDR) for the GWSB2 project in April 1991. In May 1991 WSRC Systems Engineering issued a revised Functional Design Criteria (FDC), the Rev. I document has not yet been approved by DOE. This document proposes an alternative design for the conceptual design (CDR) completed in April 1991. In June 1992 Project Management Department authorized Systems Engineering to further develop the proposed alternative design. The proposed facility will have a storage capacity for 2,268 canisters and will meet DWPF interim storage requirements for a five-year period. This document contains: a description of the proposed facility; a cost estimate of the proposed design; a cost comparison between the proposed facility and the design outlined in the FDC/CDR; and an overall assessment of the alternative design as compared with the reference FDC/CDR design

  18. Heat transfer and thermal storage performance of an open thermosyphon type thermal storage unit with tubular phase change material canisters

    International Nuclear Information System (INIS)

    Wang, Ping-Yang; Hu, Bo-Wen; Liu, Zhen-Hua

    2015-01-01

    Highlights: • A novel open heat pipe thermal storage unit is design to improve its performance. • Mechanism of its operation is phase-change heat transfer. • Tubular canisters with phase change material were placed in thermal storage unit. • Experiment and analysis are carried out to investigate its operation properties. - Abstract: A novel open thermosyphon-type thermal storage unit is presented to improve design and performance of heat pipe type thermal storage unit. In the present study, tubular canisters filled with a solid–liquid phase change material are vertically placed in the middle of the thermal storage unit. The phase change material melts at 100 °C. Water is presented as the phase-change heat transfer medium of the thermal storage unit. The tubular canister is wrapped tightly with a layer of stainless steel mesh to increase the surface wettability. The heat transfer mechanism of charging/discharging is similar to that of the thermosyphon. Heat transfer between the heat resource or cold resource and the phase change material in this device occurs in the form of a cyclic phase change of the heat-transfer medium, which occurs on the surface of the copper tubes and has an extremely high heat-transfer coefficient. A series of experiments and theoretical analyses are carried out to investigate the properties of the thermal storage unit, including power distribution, start-up performance, and temperature difference between the phase change material and the surrounding vapor. The results show that the whole system has excellent heat-storage/heat-release performance

  19. Thermal analysis of heat storage canisters for a solar dynamic, space power system

    Science.gov (United States)

    Wichner, R. P.; Solomon, A. D.; Drake, J. B.; Williams, P. T.

    1988-01-01

    A thermal analysis was performed of a thermal energy storage canister of a type suggested for use in a solar receiver for an orbiting Brayton cycle power system. Energy storage for the eclipse portion of the cycle is provided by the latent heat of a eutectic mixture of LiF and CaF2 contained in the canister. The chief motivation for the study is the prediction of vapor void effects on temperature profiles and the identification of possible differences between ground test data and projected behavior in microgravity. The first phase of this study is based on a two-dimensional, cylindrical coordinates model using an interim procedure for describing void behavor in 1-g and microgravity. The thermal analysis includes the effects of solidification front behavior, conduction in liquid/solid salt and canister materials, void growth and shrinkage, radiant heat transfer across the void, and convection in the melt due to Marangoni-induced flow and, in 1-g, flow due to density gradients. A number of significant differences between 1-g and o-g behavior were found. This resulted from differences in void location relative to the maximum heat flux and a significantly smaller effective conductance in 0-g due to the absence of gravity-induced convection.

  20. Status of Closure Welding Technology of Canister for Transportation and Storage of High Level Radioactive Material and Waste

    International Nuclear Information System (INIS)

    Lee, H. J.; Bang, K. S.; Seo, K. S.; Seo, C. S.

    2010-10-01

    Closure seal welding is one of the key technologies in fabricating and handling the canister which is used for transportation and storage of high radioactive material and waste. Simple industrial fabrication processes are used before filling the radioactive waste into the canister. But, automatic and remote processes should be used after filling the radioactive material because the thickness of canister is not sufficient to shield the high radiation from filled material or waste. In order to simplify the welding process the closure structure of canister and the sealing method are investigated and developed properly. Two types of radioactive materials such as vitrified waste and compacted solid waste are produced in nuclear industry. Because the filling method of two types of waste is different, the shapes of closure and opening of canister and welding method is also different. The canister shape and sealing method should be standardized to standardize the handling facilities and inspection process such as leak test after closure welding. In order to improve the productivity of disposal and compatibility of the canister, the structure and shape of canister should be standardized considering the type of waste. Two kind of welding process such as arc welding and resistance welding are reported and used in the field. In the arc welding process GTAW and PAW are considered proper processes for closure welding. The closure seal welding process can be selected by considering material of canister, thickness of body, productivity, and applicable codes and rules. Because the storage time of nuclear waste in canister is very long, at least 20 years, the long-time corrosion at the weld should be estimated including mechanical integrity. Recently, the mitigation of residual stress around weld region, which causes stress corrosion cracking, is also interesting research issue

  1. Performance of CASTORR HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    International Nuclear Information System (INIS)

    Wilmsmeier, Marco; Vossnacke, Andre

    2008-01-01

    On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR R HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR R HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR R HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR R HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out as well. GNS long-term CASTOR R

  2. Multi-purpose canister storage unit and transfer cask thermal analysis

    International Nuclear Information System (INIS)

    Montgomery, R.A.; Niemer, K.A.; Lindner, C.N.

    1997-01-01

    Spent Nuclear Fuel (SNF) generated at commercial nuclear power plants throughout the US is a concern because of continued delays in obtaining a safe, permanent disposal facility. Most utilities maintain their SNF in wet storage pools; however, after decades of use, many pools are filled to capacity. Unfortunately, DOE's proposed final repository at Yucca Mountain is at least 10 years from completion, and commercial power utilities have few options for SNF storage in the interim. The Multi-Purpose Canister (MPC) system, sponsored by DOE's Office of Civilian Radioactive Waste Management, is a viable solution to the interim storage problem. The system is designed for interim dry storage, transport, and ultimate disposal of commercial SNF. The MPC system consists of four separate components: an MPC, Transfer Cask, Storage Unit, and Transport Cask. The SNF assemblies are loaded and sealed inside the helium-filled steel MPC. Once sealed, the MPC is not reopened, eliminating the need to re-handle the individual spent fuel assemblies. The MPC is transferred, using the MPC Transfer Cask, into a cylindrical, reinforced-concrete Storage Unit for on-site dry storage. The MPC may be removed from the Storage Unit at any time and transferred into the MPC Transport Cask for transport to the final repository. This paper discusses the analytical approach used to evaluate the heat transfer characteristics of an MPC containing SNF assemblies in the MPC Transfer Cask and Storage Unit

  3. Fracture toughness properties of candidate canister materials for spent fuel storage by concrete cask

    International Nuclear Information System (INIS)

    Arai, Taku; Mayuzumi, Masami; Libin, Niu; Takaku, Hiroshi

    2005-01-01

    It is very significant to clarify the fracture toughness properties of candidate canister materials to ensure the structural integrity against the accidents during handling in the storage facility. Fracture toughness tests on the CT specimens cut from base metal, heat affected zone (HAZ) and weld metal in the 2 types of weld joints made by candidate canister materials (SUS329J4L duplex stainless steel and YUS270 super stainless steel) were conducted under various test temperature between 233K and 473K. Stable ductile crack extensions were observed in all of the specimens. The fracture toughness J Q of the base metal and the HAZ of SUS329L4L showed the smallest value at 233K, and increased with temperature, then reached to the largest value at 298K. At the higher temperature, the value of J Q decreased slightly with temperature. While, the value of J Q in the weld metal increased with temperature. The value of J Q of YUS270 increased with temperature. The values of J Q for weld metal in both of the materials were not greater than those in base metal and HAZ at each test temperature. The values of J Q in weld metal of both materials at 213K and 473K were greater than applied J derived from postulated semi-elliptical surface flaw and maximum allowable stress in JSME design coed. This result suggested that these materials have enough toughness for use as the canister material. (author)

  4. 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.

  5. Monitored retrievable storage and multi-purpose canister robotic applications: Feasibility, dose savings and cost analysis

    International Nuclear Information System (INIS)

    Bennett, P.C.

    1995-01-01

    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, and as an alternative to current MPC closure and welding methods at commercial nuclear reactor sites. Automation of key operational aspects is analyzed to determine equipment requirements, through-put times and equipment costs. The economic analysis approach is described, and economic and radiation dose impacts resulting from this automation are compared to manual handling methods. (author). 5 refs, 5 figs, 3 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-03

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

  7. Preliminary design of the high-level waste canister storage system: Topical report for the period of January 1, 1987--September 30, 1987

    International Nuclear Information System (INIS)

    Peters, F.E.; Leap, D.R.

    1987-11-01

    The final stage of the West Valley solidification program will be to place the high-level waste canisters in interim storage until a federal repository is ready to receive them. The waste canisters will be stored in the largest former fuel reprocessing cell at West Valley modified for this purpose. This report provides a description of the preliminary design of the Waste Canister Storage Facility. 9 refs., 14 figs., 1 tab

  8. Analysis of factors influencing the reliability of retrievable storage canisters for containment of solid high-level radioactive waste

    International Nuclear Information System (INIS)

    Mecham, W.J.; Seefeldt, W.B.; Steindler, M.J.

    1976-08-01

    The reliability of stainless steel type 304L canisters for the containment of solidified high-level radioactive wastes in the glass and calcine forms was studied. A reference system, drawn largely from information furnished by Battelle Northwest Laboratories and Atlantic Richfield Hanford Company is described. Operations include filling the canister with the appropriate waste form, interim storage at a reprocessing plant, shipment in water to a Retrievable Surface Storage Facility (RSSF), interim storage at the RSSF, and shipment to a final disposal facility. The properties of stainless steel type 304L, fission product oxides, calcine, and glass were reviewed, and mechanisms of corrosion were identified and studied. The modes of corrosion important for reliability were stress-corrosion cracking, internal pressurization of the canister by residual impurities present, intergranular attack at the waste-canister interface, and potential local effects due to migration of fission products. The key role of temperature control throughout canister lifetime is considered together with interactive effects. Methods of ameliorating adverse effects and ensuring high reliability are identified and described. Conclusions and recommendations are presented

  9. Test Plan to Determine the Maximum Surface Temperatures for a Plutonium Storage Cubicle with Horizontal 3013 Canisters

    International Nuclear Information System (INIS)

    HEARD, F.J.

    2000-01-01

    A simulated full-scale plutonium storage cubicle with 22 horizontally positioned and heated 3013 canisters is proposed to confirm the effectiveness of natural circulation. Temperature and airflow measurements will be made for different heat generation and cubicle door configurations. Comparisons will be made to computer based thermal Hydraulic models

  10. Canister Storage Building Receiving Pit Modification Informal Design Verification

    International Nuclear Information System (INIS)

    KRIEG, S.A.

    2000-01-01

    The design for modifications to the CSB Cask Receiving pit guides was verified by the informal design verification (meeting) method on August 9, 2000. The invited list of attendees and the meeting attendance sheet are included in attachment 1. The design modifications that were reviewed are documented in ECN 654484 (attachment 2). The requirement that the design is to be verified against is to ''center the transportation cask sufficiently to allow installation of the guide funnel on the cask (± 0.25 inches or less)''. The alternatives considered are detailed in attachment 3. Alternative number 4, ''Modify The Pit Guides'', was determined to be the preferred alternative primarily due to considerations of simplicity, reliability, and low cost. Alternative 1, ''Rotate the impact Absorber 180 o '', was successfully performed but was considered a temporary fix that was not acceptable for a long term operational mode. The requirement to position the receiving crane accurately enough to lower the transportation cask into the pit with the redesigned guides was discussed and considered to be achievable without undue effort from the operator. The tolerance on the OD of the transfer cask was discussed (± 1/8 inch) relative to the clearance with the guides. As-built dimensions for the cask OD will be looked at to verify sufficient clearance exists with the maximum cask OD. The final design thickness of the shims under the guides will be based on the as-built cask OD dimensions and field measurements between the pit guides. The need for a ''plastic'' cover for the guides was discussed and deemed unnecessary. Thermal growth of the cask OD was calculated at 3-5 mils and considered insignificant. The possibility of reducing the OD of the guide funnel was reviewed but this was considered impractical due to the requirement for the MCO to miss the edge of the funnel in case of a MCO drop. One of the transportation casks have the lift trunions installed 3/8 inch off center. This is not expected to present a problem, but needs to be verified. Operations personnel were not in attendance to discuss the method for lowering the cask into the pit, however it was the consensus of those who observed the lowering operation performed by startup that it is operationally viable. The proposed design will be presented to Operations for review prior to issuance of the ECN. The final conclusion of the attendees is that the design meets the requirement ''center the cask sufficiently to allow installation of the guide funnel.''

  11. Dry storage technologies: keys to choosing among metal casks, concrete shielded steel canister modules and vaults

    International Nuclear Information System (INIS)

    Roland, V.; Solignac, Y.; Chiguer, M.; Guenon, Y.

    2003-01-01

    The current international trend towards expanding Spent Fuel Interim Dry Storage capabilities goes with an improvement of the performance of the proposed systems which have to accommodate Spent fuel Assemblies characterized by ever increasing burn-up, fissile isotopes contents, thermal releases, and total inventory. Due to heterogeneous worldwide reactor pools and specific local constraints the proposed solutions have also to cope with a wide fuel design variety. Moreover, the Spent fuel Assemblies stored temporarily for cooling may have to be transported either to reprocessing facilities or to interim storage facilities before direct disposal; it is the reason why the retrievability, including or not transportability of the proposed systems, is often specified by the Utilities for the design of their Storage systems and sometimes by law. This paper shows on examples developed within companies of AREVA Group the key parameters and elements that can direct toward the selection of a technology in a user specific context. Some of the constraints are ability to dry store at once a large number of spent fuel assemblies, readily available, on a given site. No urgent need for further move of the fuel is foreseen. Then clearly a Vault Type Storage system developed and implemented by SGN is an excellent solution: It combines passive safety with immediate large capacity, which allows quick amortization of fuel receiving equipment. In addition the versatile storage position can easily accept in the same facility different fuel types, and also intermediate and High Level Waste. This is the reason why a vault system is often a preferred solution for a long-term dry interim centralized storage, for a multiplicity of spent fuel. It can be also a choice solution when the ISFSI stands on a site that is dedicated permanently to many different nuclear activities.In most cases, the producers of spent fuel require a large capacity that is cumulated over many years, each reload at a

  12. 2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

    International Nuclear Information System (INIS)

    Yaksh, M.; Wang, C.

    2004-01-01

    In the US, the number of nuclear plants expected to implement on-site dry storage is increasing each year. As reactors burn advanced fuel assemblies to higher burnups, the dry storage systems will be required to accommodate higher heat loads. This is due to the increasing capacity of the systems and the need to store higher burnup fuel with reasonable cooling periods (i.e., five to six years). As the storage systems heat rejection design must be passive, natural convection is an efficient means for rejection of heat from the spent fuel to the surface of the canister boundary. The design presented in this paper is a canistered system that employs conduction, radiation and convection to reject heat from the canister, which is stored in a vertical concrete cask. The canister containing the spent fuel in this design is a right circular stainless steel vessel capable of storing 37 PWR fuel assemblies with a total canister heat load of 40 kW. Accompanying any design effort is the use of a numerical methodology that can accurately predict the peak-clad temperatures of the fuel and the structural components of the system. The main challenge to any analysis employing internal natural convection may be perceived as a practical limitation due to the size of the model. Since canisters are typically cylindrical, a two-dimensional model can be used to represent the canister. The fuel basket structure, which maintains the configuration of the spent fuel, is an array of square tubes, and is non-axisymmetric. Flow up through the fuel region in the basket encounters a complex cross section due to the fuel assembly rod array (up to 17 x 17). The flow region of the heated gas down the outside of the basket in the annulus between the canister shell and the basket assembly (downcomer) is also an irregular shaped area. To confirm that a two-dimensional (2D) modelling methodology is appropriate, a benchmark using results from a thermal test is required. The thermal test focuses on the

  13. CFD Analysis on the Passive Heat Removal by Helium and Air in the Canister of Spent Fuel Dry Storage System

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Do Young; Jeong, Ui Ju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    In the current commercial design, the canister of the dry storage system is mainly backfilled with helium gas. Helium gas shows very conductive behavior due to high thermal conductivity and small density change with temperature. However, other gases such as air, argon, or nitrogen are expected to show effective convective behavior. Thus these are also considered as candidates for the backfill gas to provide effective coolability. In this study, to compare the dominant cooling mechanism and effectiveness of cooling between helium gas and air, a computational fluid dynamics (CFD) analysis for the canister of spent fuel dry storage system with backfill gas of helium and air is carried out. In this study, CFD simulations for the helium and air backfilled gas for dry storage system canister were carried out using ANSYS FLUENT code. For the comparison work, two backfilled fluids were modeled with same initial and boundary conditions. The observed major difference can be summarized as follows. - The simulation results showed the difference in dominant heat removal mechanism. Conduction for helium, and convection for air considering Reynolds number distribution. - The temperature gradient inside the fuel assembly showed that in case of air, more effective heat mixing occurred compared to helium.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-24

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

  15. Performance of CASTOR {sup registered} HAW cask cold trials for loading, Transport and storage of HAW canisters

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsmeier, Marco; Horn, Thomas; Graf, Wilhelm [GNS Gesellschaft fuer Nuklear-Service mbH (Germany)

    2009-07-01

    With over 30 years of experience in the design, manufacturing, assembly and loading of CASTOR {sup registered} casks, GNS is one of the worldwide leading suppliers of casks for the transport and storage of spent fuel assemblies as well as for canisters with vitrified high active wastes (meanwhile over 1.000 casks loaded and stored and more than 1.500 ordered). GNS's products are used at around 30 sites worldwide for a wide range of inventories from pressurised and boiling water reactor fuels (PWR, VVER and BWR, RBMK), thorium high-temperature reactor fuels (THTR) and research reactor fuels (MTR) to vitrified high active wastes (HAW) from reprocessing plants. GNS is responsible for all nuclear wastes resulting from German Nuclear Power Plants and assists and/or performs in the loading and dispatch of CASTOR {sup registered} casks as well as their transport to and storage at central interim storage facilities and local interim storage areas. (orig.)

  16. Performance of CASTOR{sup R} HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsmeier, Marco; Vossnacke, Andre [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany)

    2008-07-01

    On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR{sup R} HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR{sup R} HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR{sup R} HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR{sup R} HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out

  17. The concrete canister program

    International Nuclear Information System (INIS)

    Ohta, M.M.

    1978-02-01

    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)

  18. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    B. Gorpani

    2000-01-01

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

  19. Canister Transfer System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

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

  20. Measures of stress corrosion cracking in the canister storage facility of spent nuclear fuel. Vol.3. Development of salt particle collection device

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Saegusa, Toshiari

    2009-01-01

    A natural ventilation system is generally adopted for storage facilities of spent nuclear fuel. At the storage facilities of concrete casks built near the seashore, the air including the sea salt particles comes into the concrete casks and could cause SCC to the canister made of stainless steel. In this study, we proposed a salt particle collection device with a low flow resistance which does not block the air flow into the building. The effect of the device was evaluated quantitatively in laboratory experiments and in field tests. Obtained results are as follows: (1) The pressure loss of the device is smaller than one-sevenths of pressure loss of a filter used in a forced ventilation system and the efficiency of salt particle collection is more than 80% in both laboratory experiments and field tests. However, the efficiency of salt particle collection depends on the diameter of a salt particle. (2) It was clarified the diameter of the particle which can be collected by the device under the condition of the size of the device, the density and velocity of the particle. And the pressure loss of the device was evaluated. In the case of setting the device in the air inlet of a concrete cask, salt particles lager than 27μm in diameter can be collected by the device under the condition of the same pressure loss of a bard screen which opening ratio is 80%. (author)

  1. Analysis of Samples Collected from the Surface of Interim Storage Canisters at Calvert Cliffs in June 2017: Revision 01.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Schindelholz, Eric John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-11-01

    In June 2017, dust and salt samples were collected from the surface of Spent Nuclear Fuel (SNF) dry storage canisters at the Calvert Cliffs Nuclear Power Plant. The samples were delivered to Sandia National laboratories for analysis. Two types of samples were collected: filter-backed Scotch-Brite TM pads were used to collect dry dust samples for characterization of salt and dust morphologies and distributions; and Saltsmart TM test strips were used to collect soluble salts for determining salt surface loadings per unit area. After collection, the samples were sealed into plastic sleeves for shipping. Condensation within the sleeves containing the Scotch-Brite TM samples remobilized the salts, rendering them ineffective for the intended purpose, and also led to mold growth, further compromising the samples; for these reasons, the samples were not analyzed. The SaltSmart TM samples were unaffected and were analyzed by ion chromatography for major anions and cations. The results of those analyses are presented here.

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

    International Nuclear Information System (INIS)

    Bond, A.E.; Hoch, A.R.; Jones, G.D.; Tomczyk, A.J.; Wiggin, R.M.; Worraker, W.J.

    1997-05-01

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

  3. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    Nilsson, Fred

    1992-12-01

    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

  4. Energy in buildings: Efficiency, renewables and storage

    Science.gov (United States)

    Koebel, Matthias M.

    2017-07-01

    This lecture summary provides a short but comprehensive overview on the "energy and buildings" topic. Buildings account for roughly 40% of the global energy demands. Thus, an increased adoption of existing and upcoming materials and solutions for the building sector represents an enormous potential to reduce building related energy demands and greenhouse gas emissions. The central question is how the building envelope (insulation, fenestration, construction style, solar control) affects building energy demands. Compared to conventional insulation materials, superinsulation materials such as vacuum insulation panels and silica aerogel achieve the same thermal performance with significantly thinner insulation layers. With low-emissivity coatings and appropriate filler gasses, double and triple glazing reduce thermal losses by up to an order of magnitude compared to old single pane windows, while vacuum insulation and aerogel filled glazing could reduce these even further. Electrochromic and other switchable glazing solutions maximize solar gains during wintertime and minimize illumination demands whilst avoiding overheating in summer. Upon integration of renewable energy systems into the building energy supply, buildings can become both producers and consumers of energy. Combined with dynamic user behavior, temporal variations in the production of renewable energy require appropriate storage solutions, both thermal and electrical, and the integration of buildings into smart grids and energy district networks. The combination of these measures allows a reduction of the existing building stock by roughly a factor of three —a promising, but cost intensive way, to prepare our buildings for the energy turnaround.

  5. A preliminary study of a miniature planar 6-cell PEMFC stack combined with a small hydrogen storage canister

    Science.gov (United States)

    Zhang, Xigui; Zheng, Dan; Wang, Tao; Chen, Cong; Cao, Jianyu; Yan, Jian; Wang, Wenming; Liu, Juanying; Liu, Haohan; Tian, Juan; Li, Xinxin; Yang, Hui; Xia, Baojia

    The fabrication and performance evaluation of a miniature 6-cell PEMFC stack based on Micro-Electronic-Mechanical-System (MEMS) technology is presented in this paper. The stack with a planar configuration consists of 6-cells in serial interconnection by spot welding one cell anode with another cell cathode. Each cell was made by sandwiching a membrane-electrode-assembly (MEA) between two flow field plates fabricated by a classical MEMS wet etching method using silicon wafer as the original material. The plates were made electrically conductive by sputtering a Ti/Pt/Au composite metal layer on their surfaces. The 6-cells lie in the same plane with a fuel buffer/distributor as their support, which was fabricated by the MEMS silicon-glass bonding technology. A small hydrogen storage canister was used as fuel source. Operating on dry H 2 at a 40 ml min -1 flow rate and air-breathing conditions at room temperature and atmospheric pressure, the linear polarization experiment gave a measured peak power of 0.9 W at 250 mA cm -2 for the stack and average power density of 104 mW cm -2 for each cell. The results suggested that the stack has reasonable performance benefiting from an even fuel supply. But its performance tended to deteriorate with power increase, which became obvious at 600 mW. This suggests that the stack may need some power assistance, from say supercapacitors to maintain its stability when operated at higher power.

  6. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  7. Remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility is being constructed at the Savannah River Plant at a cost of $870 million to immobilize the defense high-level radioactive waste. This radioactive waste is being added to borosilicate glass for later disposal in a federal repository. The borosilicate glass is poured into stainless steel canisters for storage. These canisters must be handled remotely because of their high radioactivity, up to 5000 R/h. After the glass has been poured into the canister which will be temporarily sealed, it is transferred to a decontamination cell and decontaminated. The canister is then transferred to the weld cell where a permanent cap is welded into place. The canisters must then be transported from the processing building to a storage vault on the plant until the federal repository is available. A shielded canister transporter (SCT) has been designed and constructed for this purpose. The design of the SCT vehicle allows the safe transport of a highly radioactive canister containing borosilicate glass weighing 2300 kg with a radiation level up to 5000 R/h from one building to another. The design provides shielding for the operator in the cab of the vehicle to be below 0.5 rem/h. The SCT may also be used to load the final shipping cask when the federal repository is ready to receive the canisters

  8. Efficient storage mechanisms for building better supercapacitors

    Science.gov (United States)

    Salanne, M.; Rotenberg, B.; Naoi, K.; Kaneko, K.; Taberna, P.-L.; Grey, C. P.; Dunn, B.; Simon, P.

    2016-06-01

    Supercapacitors are electrochemical energy storage devices that operate on the simple mechanism of adsorption of ions from an electrolyte on a high-surface-area electrode. Over the past decade, the performance of supercapacitors has greatly improved, as electrode materials have been tuned at the nanoscale and electrolytes have gained an active role, enabling more efficient storage mechanisms. In porous carbon materials with subnanometre pores, the desolvation of the ions leads to surprisingly high capacitances. Oxide materials store charge by surface redox reactions, leading to the pseudocapacitive effect. Understanding the physical mechanisms underlying charge storage in these materials is important for further development of supercapacitors. Here we review recent progress, from both in situ experiments and advanced simulation techniques, in understanding the charge storage mechanism in carbon- and oxide-based supercapacitors. We also discuss the challenges that still need to be addressed for building better supercapacitors.

  9. Remote automatic plasma arc-closure welding of a dry-storage canister for spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Sprecace, R.P.; Blankenship, W.P.

    1982-01-01

    A carbon steel storage canister has been designed for the dry encapsulation of spent nuclear fuel assemblies or of logs of vitrified high level radioactive waste. The canister design is in conformance with the requirements of the ASME Code, Section III, Division 1 for a Class 3 vessel. The canisters will be loaded and sealed as part of a completely remote process sequence to be performed in the hot bay of an experimental encapsulation facility at the Nevada Test Site. The final closure to be made is a full penetration butt weld between the canister body, a 12.75-in O.D. x 0.25-in wall pipe, and a mating semiellipsoidal closure lid. Due to a combination of design, application and facility constraints, the closure weld must be made in the 2G position (canister vertical). The plasma arc welding system is described, and the final welding procedure is described and discussed in detail. Several aspects and results of the procedure development activity, which are of both specific and general interest, are highlighted; these include: The critical welding torch features which must be exactly controlled to permit reproducible energy input to, and gas stream interaction with, the weld puddle. A comparison of results using automatic arc voltage control with those obtained using a mechanically fixed initial arc gap. The optimization of a keyhole initiation procedure. A comparison of results using an autogenous keyhole closure procedure with those obtained using a filler metal addition. The sensitivity of the welding process and procedure to variations in joint configuration and dimensions and to variations in base metal chemistry. Finally, the advantages and disadvantages of the plasma arc process for this application are summarized from the current viewpoint, and the applicability of this process to other similar applications is briefly indicated

  10. Transport of multiassembly sealed canisters

    International Nuclear Information System (INIS)

    Quinn, R.D.; Lehnert, R.A.; Rosa, J.M.

    1992-01-01

    A significant portion of the commercial spent nuclear fuel in dry storage in the US will be stored in multiassembly sealed canisters before the DOE begins accepting fuel from utilities in 1998. This paper reports that it is desirable from economic and ALARA perspectives to transfer these canisters directly from the plant to the MRS. To this end, it is necessary that the multiassembly sealed canisters, which have been licensed for storage under 10CFR72, be qualified for shipment within a suitable shipping cask under the rules of 10CFR71. Preliminary work performed to date indicates that it is feasible to license a current canister design for transportation, and work is proceeding on obtaining NRC approval

  11. The Role of Energy Storage in Commercial Building

    Energy Technology Data Exchange (ETDEWEB)

    Kintner-Meyer, Michael CW; Subbarao, Krishnappa; Prakash Kumar, Nirupama; Bandyopadhyay, Gopal K.; Finley, C.; Koritarov, V. S.; Molburg, J. C.; Wang, J.; Zhao, Fuli; Brackney, L.; Florita, A. R.

    2010-09-30

    Motivation and Background of Study This project was motivated by the need to understand the full value of energy storage (thermal and electric energy storage) in commercial buildings, the opportunity of benefits for building operations and the potential interactions between a building and a smart grid infrastructure. On-site or local energy storage systems are not new to the commercial building sector; they have been in place in US buildings for decades. Most building-scale storage technologies are based on thermal or electrochemical storage mechanisms. Energy storage technologies are not designed to conserve energy, and losses associated with energy conversion are inevitable. Instead, storage provides flexibility to manage load in a building or to balance load and generation in the power grid. From the building owner's perspective, storage enables load shifting to optimize energy costs while maintaining comfort. From a grid operations perspective, building storage at scale could provide additional flexibility to grid operators in managing the generation variability from intermittent renewable energy resources (wind and solar). To characterize the set of benefits, technical opportunities and challenges, and potential economic values of storage in a commercial building from both the building operation's and the grid operation's view-points is the key point of this project. The research effort was initiated in early 2010 involving Argonne National Laboratory (ANL), the National Renewable Energy Laboratory (NREL), and Pacific Northwest National Laboratory (PNNL) to quantify these opportunities from a commercial buildings perspective. This report summarizes the early discussions, literature reviews, stakeholder engagements, and initial results of analyses related to the overall role of energy storage in commercial buildings. Beyond the summary of roughly eight months of effort by the laboratories, the report attempts to substantiate the importance of

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

    Energy Technology Data Exchange (ETDEWEB)

    Wiborg, J.C.

    1995-12-01

    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.

  13. Shippingport Spent Fuel Canister System Description

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    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. Sampling and analysis plan for sludge located in fuel storage canisters of the 105-K West basin

    International Nuclear Information System (INIS)

    Baker, R.B.

    1997-01-01

    This Sampling and Analysis Plan (SAP) provides direction for the first sampling of sludge from the K West Basin spent fuel canisters. The specially developed sampling equipment removes representative samples of sludge while maintaining the radioactive sample underwater in the basin pool (equipment is described in WHC-SD-SNF-SDD-004). Included are the basic background logic for sample selection, the overall laboratory analyses required and the laboratory reporting required. These are based on requirements put forth in the data quality objectives (WHC-SD-SNF-DQO-012) established for this sampling and characterization activity

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

    International Nuclear Information System (INIS)

    BLACK, D.M.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    BLACK, D.M.

    1999-01-01

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

  17. Multi-Canister Overpack (MCO) Topical Report

    International Nuclear Information System (INIS)

    LORENZ, B.D.

    2000-01-01

    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

  18. Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes

    International Nuclear Information System (INIS)

    Harmon, K.M.; Johnson, A.B. Jr.

    1984-04-01

    The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage

  19. Research of radioactive waste storage cask/canister materials, spent nuclear fuels and various radioactive waste forms and development of their assessment methods. Final report for Stage 3

    International Nuclear Information System (INIS)

    Dobrev, D.; Balek, V.; Červinka, R.; Večerník, P.; Člupek, M.; Kouřil, M.; Novák, P.; Stoulil, J.; Silber, R.

    2013-08-01

    The main topics treated are: Research and development of methodologies for canister/cask material degradation assessment; Laboratory research of selected materials of canister/cask with radioactive waste; and Research and assessment of canister/cask materials in natural granite rocks. Two additional documents are appended: Corrosion rate determination for samples in compacted bentonite in anaerobic conditions (methodology), and Roll test for corrosion test in an occluded solution at the interface between a radioactive waste disposal canister and the bentonite cover. (P.A.)

  20. Unreviewed safety question evaluation of 100 K West fuel canister gas and liquid sampling

    International Nuclear Information System (INIS)

    Alwardt, L.D.

    1995-01-01

    The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation for the gas and liquid sampling activities associated with the fuel characterization program at the 100 K West (KW) fuel storage basin. The scope of this safety evaluation is limited to the movement of canisters between the main storage basin, weasel pit, and south loadout pit transfer channel (also known as the decapping station); gas and liquid sampling of fuel canisters in the weasel pit; mobile laboratory preliminary sample analysis in or near the 105 KW basin building; and the placement of sample containers in an approved shipping container. It was concluded that the activities and potential accident consequences associated with the gas and liquid sampling of 100 KW fuel canisters are bounded by the current safety basis documents and do not constitute an Unreviewed Safety Question

  1. Air quality in low-ventilated museum storage buildings

    DEFF Research Database (Denmark)

    Ryhl-Svendsen, Morten; Aasbjerg Jensen, Lars; Klenz Larsen, Poul

    2014-01-01

    Modern low-energy museum storage buildings are often designed for a low air exchange rate, on the order of less than 1 exchange per day. We investigated how this affected the indoor air quality in six Danish museum storage buildings. The infiltration of ambient pollutants, and the level to which...... internally-generated pollutants accumulate, were measured by passive sampling of ozone, nitrogen dioxide, and organic acids. The air exchange rates and the interchange of air between storage rooms were measured by the per-fluorocarbon tracer gas method. Ambient pollutants were reduced in concentration...

  2. Spent fuel consolidation in the 105KW Building fuel storage basin

    International Nuclear Information System (INIS)

    Johnson, B.H.

    1994-01-01

    This study is one element of a larger engineering study effort by WHC to examine the feasibility of irradiated fuel and sludge consolidation in the KW Basin in response to TPA Milestone (target date) M-34-00-T03. The study concludes that up to 11,500 fuel storage canisters could be accommodated in the KW Basin with modifications. These modifications would include provisions for multi-tiered canister storage involving the fabrication and installation of new storage racks and installation of additional decay heat removal systems for control of basin water temperature. The ability of existing systems to control radionuclide concentrations in the basin water is examined. The study discusses requirements for spent nuclear fuel inventory given the proposed multi-tiered storage arrangement, the impact of the consolidated mass on the KW Basin structure, and criticality issues associated with multi-tiered storage

  3. Solar Heating System with Building-Integrated Heat Storage

    DEFF Research Database (Denmark)

    Heller, Alfred

    1996-01-01

    Traditional solar heating systems cover between 5 and 10% of the heat demand fordomestic hot water and comfort heating. By applying storage capacity this share can beincreased much. The Danish producer of solar heating systems, Aidt-Miljø, markets such a system including storage of dry sand heated...... by PP-pipe heat exchanger. Heat demand is reduced due to direct solar heating, and due to storage. Heat demand is reduced due to direct solar heating, due to storage and due to lower heat losses through the ground. In theory, by running the system flow backwards through the sand storage, active heating...... can be achieved.The objective of the report is to present results from measured system evaluation andcalculations and to give guidelines for the design of such solar heating systems with building integrated sand storage. The report is aimed to non-technicians. In another report R-006 the main results...

  4. Environmental permits and approvals plan for high-level waste interim storage, Project W-464

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1998-01-01

    This report discusses the Permitting Plan regarding NEPA, SEPA, RCRA, and other regulatory standards and alternatives, for planning the environmental permitting of the Canister Storage Building, Project W-464

  5. Passive hygrothermal control of a museum storage building in Vejle

    DEFF Research Database (Denmark)

    Christensen, Jørgen Erik; Janssen, Hans

    2010-01-01

    and maintenance costs are currently motivating a paradigm change toward passive control. Passive control, via the thermal and hygric inertia of the building, is gaining a foothold in the museum conservation and building physical community. In this report we document the hygrothermal performance optimisation...... of a museum storage building, related to an existing storage centre in Vejle (Denmark). The current building design already incorporates passive control concepts: thermal inertia is provided by the thick walls, the ground floor and its underlying soil volume, while hygric inertia is provided by the thick...... with the thermal inertia of the ground and thus a higher temperature variation; For those reasons, more heavily insulated walls and roofs could be considered. Their effects on the interior climate and dehumidification load are however not that large. For the floor, no insulation should be added, and it could...

  6. Building a mass storage system for physics applications

    International Nuclear Information System (INIS)

    Holmes, H.; Loken, S.

    1991-03-01

    The IEEE Mass Storage Reference Model and forthcoming standards based on it provide a standardized architecture to facilitate designing and building mass storage systems, and standard interfaces so that hardware and software from different vendors can interoperate in providing mass storage capabilities. A key concept of this architecture is the separation of control and data flows. This separation allows a smaller machine to provide control functions, while the data can flow directly between high-performance channels. Another key concept is the layering of the file system and the storage functions. This layering allows the designers of the mass storage system to focus on storage functions, which can support a variety of file systems, such as the Network File System, the Andrew File System, and others. The mass storage system provides location-independent file naming, essential if files are to be migrated to different storage devices without requiring changes in application programs. Physics data analysis applications are particularly challenging for mass storage systems because they stream vast amounts of data through analysis applications. Special mechanisms are required, to handle the high data rates and to avoid upsetting the caching mechanisms commonly used for smaller, repetitive-use files. High data rates are facilitated by direct channel connections, where, for example, a dual-ported drive will be positioned by the mass storage controller on one channel, then the data will flow on a second channel directly into the user machine, or directly to a high capacity network, greatly reducing the I/O capacity required in the mass storage control computer. Intelligent storage allocation can be used to bypass the cache devices entirely when large files are being moved

  7. Building with integral solar-heat storage--Starkville, Mississippi

    Science.gov (United States)

    1981-01-01

    Column supporting roof also houses rock-storage bin of solar-energy system supplying more than half building space heating load. Conventional heaters supply hot water. Since bin is deeper and narrower than normal, individual pebble size was increased to keep airflow resistance at minimum.

  8. Passive hygrothermal control of a museum storage building

    DEFF Research Database (Denmark)

    Christensen, Jørgen Erik; Janssen, Hans

    2011-01-01

    of C02 emission. The purpose for this paper is to show that it is possible to reach the goal of using renewable energy for museum storage buildings by rethinking the strategy for the dehumidification design and in this way contribute to a C02 neutral environment. The solution is to construct a very...

  9. Canister disposition plan for the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.; Payne, C.H.

    1990-01-01

    This report details the disposition of canisters and the canistered waste forms produced during the DWPF Startup Test Program. The six melter campaigns (DWPF Startup Tests FA-13, WP-14, WP-15, WP-16, WP-17, and FA-18) will produce 126 canistered waste forms. In addition, up to 20 additional canistered waste forms may be produced from glass poured during the transition between campaigns. In particular, this canister disposition plan (1) assigns (by alpha-numeric code) a specific canister to each location in the six campaign sequences, (2) describes the method of access for glass sampling on each canistered waste form, (3) describes the nature of the specific tests which will be carried out, (4) details which tests will be carried out on each canistered waste form, (5) provides the sequence of these tests for each canistered waste form, and (6) assigns a storage location for each canistered waste form. The tests are designed to provide evidence, as detailed in the Waste Form Compliance Plan (WCP 1 ), that the DWPF product will comply with the Waste Acceptance Product Specifications (WAPS 2 ). The WAPS must be met before the canistered waste form is accepted by DOE for ultimate disposal at the Federal Repository. The results of these tests will be included in the Waste Form Qualification Report (WQR)

  10. Peak reduction for commercial buildings using energy storage

    Science.gov (United States)

    Chua, K. H.; Lim, Y. S.; Morris, S.

    2017-11-01

    Battery-based energy storage has emerged as a cost-effective solution for peak reduction due to the decrement of battery’s price. In this study, a battery-based energy storage system is developed and implemented to achieve an optimal peak reduction for commercial customers with the limited energy capacity of the energy storage. The energy storage system is formed by three bi-directional power converter rated at 5 kVA and a battery bank with capacity of 64 kWh. Three control algorithms, namely fixed-threshold, adaptive-threshold, and fuzzy-based control algorithms have been developed and implemented into the energy storage system in a campus building. The control algorithms are evaluated and compared under different load conditions. The overall experimental results show that the fuzzy-based controller is the most effective algorithm among the three controllers in peak reduction. The fuzzy-based control algorithm is capable of incorporating a priori qualitative knowledge and expertise about the load characteristic of the buildings as well as the useable energy without over-discharging the batteries.

  11. Multi-purpose canister project overview

    International Nuclear Information System (INIS)

    Williams, J.

    1995-01-01

    In this presentation, the author lists the approved and proposed dry storage technologies. He discusses the compatibility of dry storage systems with waste management systems. Historical aspects, recent history, key features of the program approach, benefits, specifications, acquisition and potential utility use of the multi-purpose canister (MPC) are covered. The MPCs provide standardization in the waste management system and a cost savings to utilities and government. MPC will be developed to the same level as existing dry storage systems

  12. SRL canister impact tests

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-05-01

    The Defense Waste Processing Facility (DWPF) is being constructed at the SRP for the containerization of high-level nuclear waste as a waste form for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in Type 304L stainless steel canisters 2 feet in diameter x 9 feet 10 inches long. The canisters have a minimum wall thickness of 3/8 inch. Over a three-year period, nineteen drop-tests of nine canisters, filled with simulated waste glass, were made in support of the DWPF containerization program. Eight of the canister evaluation tests were of Type 304L stainless steel material and one was of commercially pure titanium. Three different length (9.44, 5.06, and 7.88 inch) nozzle configurations containing final closure upset welds were evaluated for the stainless steel canisters. All impact tests of the stainless steel canisters, which included bottom-, side-, and top-drops, were acceptable. The bottom-drop test of the titanium canister, which contained a final closure upset weld, was acceptable; however, the top-drop resulted in a breaching of the top head where it joins the nozzle. The final closure titanium upset weld was acceptable. The titanium canister wall thickness was 1/4 inch

  13. 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.

  14. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1993-01-01

    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

  15. Thermal energy storage for cooling of commercial buildings

    Energy Technology Data Exchange (ETDEWEB)

    Akbari, H. (Lawrence Berkeley Lab., CA (USA)); Mertol, A. (Science Applications International Corp., Los Altos, CA (USA))

    1988-07-01

    The storage of coolness'' has been in use in limited applications for more than a half century. Recently, because of high electricity costs during utilities' peak power periods, thermal storage for cooling has become a prime target for load management strategies. Systems with cool storage shift all or part of the electricity requirement from peak to off-peak hours to take advantage of reduced demand charges and/or off-peak rates. Thermal storage technology applies equally to industrial, commercial, and residential sectors. In the industrial sector, because of the lack of economic incentives and the custom design required for each application, the penetration of this technology has been limited to a few industries. The penetration rate in the residential sector has been also very limited due to the absence of economic incentives, sizing problems, and the lack of compact packaged systems. To date, the most promising applications of these systems, therefore, appear to be for commercial cooling. In this report, the current and potential use of thermal energy storage systems for cooling commercial buildings is investigated. In addition, a general overview of the technology is presented and the applicability and cost-effectiveness of this technology for developed and developing countries are discussed. 28 refs., 12 figs., 1 tab.

  16. Integrated building energy systems design considering storage technologies

    Energy Technology Data Exchange (ETDEWEB)

    Stadler, Michael; Marnay, Chris; Lai, Judy; Aki, Hirohisa (Lawrence Berkeley National Laboratory (United States)). e-mail: MStadler@lbl.gov; Siddiqui, Afzal (Dept. of Statistical Science at Univ. College London (United Kingdom))

    2009-07-01

    The addition of storage technologies such as flow batteries, conventional batteries, and heat storage can improve the economic, as well as environmental attraction of micro-generation systems (e.g. PV or fuel cells with or without CHP) and contribute to enhanced demand response. The interactions among PV, solar thermal, and storage systems can be complex, depending on the tariff structure, load profile, etc. In order to examine the impact of storage technologies on demand response and CO{sub 2} emissions, a microgrid's distributed energy resources (DER) adoption problem is formulated as a mixed-integer linear program that can pursue two strategies as its objective function. These two strategies are minimization of its annual energy costs or of its CO{sub 2} emissions. The problem is solved for a given test year at representative customer sites, e.g. nursing homes, to obtain not only the optimal investment portfolio, but also the optimal hourly operating schedules for the selected technologies. This paper focuses on analysis of storage technologies in micro-generation optimization on a building level, with example applications in New York State and California. It shows results from a two-year research project performed for the U.S. Dept. of Energy and ongoing work. Contrary to established expectations, our results indicate that PV and electric storage adoption compete rather than supplement each other considering the tariff structure and costs of electricity supply. The work shows that high electricity tariffs during on-peak hours are a significant driver for the adoption of electric storage technologies. To satisfy the site's objective of minimizing energy costs, the batteries have to be charged by grid power during off-peak hours instead of PV during on-peak hours. In contrast, we also show a CO{sub 2} minimization strategy where the common assumption that batteries can be charged by PV can be fulfilled at extraordinarily high energy costs for the site.

  17. Integrated Building Energy Systems Design Considering Storage Technologies

    International Nuclear Information System (INIS)

    Stadler, Michael; Marnay, Chris; Siddiqui, Afzal; Lai, Judy; Aki, Hirohisa

    2009-01-01

    The addition of storage technologies such as flow batteries, conventional batteries, and heat storage can improve the economic, as well as environmental attraction of micro-generation systems (e.g., PV or fuel cells with or without CHP) and contribute to enhanced demand response. The interactions among PV, solar thermal, and storage systems can be complex, depending on the tariff structure, load profile, etc. In order to examine the impact of storage technologies on demand response and CO2 emissions, a microgrid's distributed energy resources (DER) adoption problem is formulated as a mixed-integer linear program that can pursue two strategies as its objective function. These two strategies are minimization of its annual energy costs or of its CO2 emissions. The problem is solved for a given test year at representative customer sites, e.g., nursing homes, to obtain not only the optimal investment portfolio, but also the optimal hourly operating schedules for the selected technologies. This paper focuses on analysis of storage technologies in micro-generation optimization on a building level, with example applications in New York State and California. It shows results from a two-year research project performed for the U.S. Department of Energy and ongoing work. Contrary to established expectations, our results indicate that PV and electric storage adoption compete rather than supplement each other considering the tariff structure and costs of electricity supply. The work shows that high electricity tariffs during on-peak hours are a significant driver for the adoption of electric storage technologies. To satisfy the site's objective of minimizing energy costs, the batteries have to be charged by grid power during off-peak hours instead of PV during on-peak hours. In contrast, we also show a CO2 minimization strategy where the common assumption that batteries can be charged by PV can be fulfilled at extraordinarily high energy costs for the site

  18. The vibration measurements at the photon factory storage ring building

    International Nuclear Information System (INIS)

    Haga, K.; Nakayama, M.; Masuda, K.; Ishizaki, H.; Kura, M.; Meng, L.; Oku, Y.

    1999-01-01

    The Photon Factory is a 2.5 GeV electron storage ring and has been operating since 1982 as a dedicated SR source. At the Photon Factory, we have been pursuing the various sources of the beam instabilities which deteriorated the SR beam quality in the wide frequency range. Some of the sources were the vibrations of magnets and floor of the ring tunnel, temperature change of the cooling water and the elongation of the storage ring building roof due to sunshine that induced the diurnal motion of the SR beam axis. This article presents the results of the vibration measurements that have been performed at the Photon Factory storage ring building. (1) The vibrations of the ring tunnel floor and the experimental hall floor, comparing with the vibration of the ground surrounding the storage ring building, are same order in the 1 ∼ 5 Hz range, and 1/3 ∼ 1/5 in the 5 ∼ 100 Hz range, in the vertical and the horizontal direction. (2) The effects of the vibration arising from the operating eight air-conditioners can be seen in the Fourier spectrum of the vibration of the ring tunnel floor, experimental floor, Q-magnets and BPM vacuum duct. (3) The vibrations of the Q-magnet and girder at frequencies near their fundamental resonant frequencies have been amplified 100 limes in the lateral direction comparing to the floor vibration. (4) Correlation between the vibration of the BPM vacuum duct and the vibration of the electron beam motion is unknown for the lack of the precise data. (authors)

  19. Final Safety Analysis Document for Building 693 Chemical Waste Storage Building at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    Salazar, R.J.; Lane, S.

    1992-02-01

    This Safety Analysis Document (SAD) for the Lawrence Livermore National Laboratory (LLNL) Building 693, Chemical Waste Storage Building (desipated as Building 693 Container Storage Unit in the Laboratory's RCRA Part B permit application), provides the necessary information and analyses to conclude that Building 693 can be operated at low risk without unduly endangering the safety of the building operating personnel or adversely affecting the public or the environment. This Building 693 SAD consists of eight sections and supporting appendices. Section 1 presents a summary of the facility designs and operations and Section 2 summarizes the safety analysis method and results. Section 3 describes the site, the facility desip, operations and management structure. Sections 4 and 5 present the safety analysis and operational safety requirements (OSRs). Section 6 reviews Hazardous Waste Management's (HWM) Quality Assurance (QA) program. Section 7 lists the references and background material used in the preparation of this report Section 8 lists acronyms, abbreviations and symbols. Appendices contain supporting analyses, definitions, and descriptions that are referenced in the body of this report

  20. Staging and storage facility feasibility study. Final report

    International Nuclear Information System (INIS)

    Swenson, C.E.

    1995-02-01

    This study was performed to investigate the feasibility of adapting the design of the HWVP Canister Storage Building (CSB) to meet the needs of the WHC Spent Nuclear Fuel Project for Staging and Storage Facility (SSF), and to develop Rough Order of Magnitude (ROM) cost and schedule estimates

  1. TMI-2 fuel canister interface requirements for INEL. Revision 1

    International Nuclear Information System (INIS)

    Wilkins, D.E.; Martz, D.E.; Reno, H.W.

    1984-06-01

    This report focuses on fuel canister interface requirements at INEL which should be incorporated into the canister design criteria. The requirements will ensure compatibility with existing INEL structures and equipment to be used for receipt, unloading, and storage of fuel canisters. INEL can and does receive and store radioactive materials in many different forms, including reactor fuel. INEL requires detailed descriptions of canisters and casks. Therefore, requirements listed represent engineering design features which will simplify the handling and storage operations; consequently, they are not to be viewed as absolute or non-negotiable. However, the core acquisition contract was negotiated with certain storage assumptions which effect costs of storage. Deviations from those assumptions which significantly effect costs would require approval by DOE-Idaho. If some stated requirements are too restrictive, modifications based on sound engineering principles may be negotiated with INEL. 11 figures

  2. 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.

  3. Manufacture of disposal canisters

    International Nuclear Information System (INIS)

    Nolvi, L.

    2009-12-01

    The report summarizes the development work carried out in the manufacturing of disposal canister components, and present status, in readiness for manufacturing, of the components for use in assembly of spent nuclear fuel disposal canister. The disposal canister consist of two major components: the nodular graphite cast iron insert and overpack of oxygen-free copper. The manufacturing process for copper components begins with a cylindrical cast copper billet. Three different manufacturing processes i.e. pierce and draw, extrusion and forging are being developed, which produce a seamless copper tube or a tube with an integrated bottom. The pierce and draw process, Posiva's reference method, makes an integrated bottom possible and only the lid requires welding. Inserts for BWR-element are cast with 12 square channels and inserts for VVER 440-element with 12 round channels. Inserts for EPR-elements have four square channels. Casting of BWR insert type has been studied so far. Experience of casting inserts for PWR, which is similar to the EPR-type, has been got in co-operation with SKB. The report describes the processes being developed for manufacture of disposal canister components and some results of the manufacturing experiments are presented. Quality assurance and quality control in manufacture of canister component is described. (orig.)

  4. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    McCoy, H.E.

    1983-01-01

    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 300 0 C, 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 300 0 C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800 0 C 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

  5. Advanced storage concepts for solar and low energy buildings, IEA-SHC Task 32. Slutrapport

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, J.M.; Andersen, Elsa; Furbo, S.

    2008-01-15

    This report reports on the results of the activities carried through in connection with the Danish part of the IEA SHC Task 32 project: Advanced Storage Concepts for Solar and Low Energy Buildings. The Danish involvement has focused on Subtask C: Storage Concepts Based on Phase Change Materials and Subtask D: Storage Concepts Based on Advanced Water Tanks and Special Devices. The report describes activities concerning heat-of-fusion storage and advanced water storage. (BA)

  6. A Film Canister Colorimeter.

    Science.gov (United States)

    Gordon, James; James, Alan; Harman, Stephanie; Weiss, Kristen

    2002-01-01

    A low-cost, low-tech colorimeter was constructed from a film canister. The student-constructed colorimeter was used to show the Beer-Lambert relationship between absorbance and concentration and to calculate the value of the molar absorptivity for permanganate at the wavelength emission maximum for an LED. Makes comparisons between this instrument…

  7. Central unresolved issues in thermal energy storage for building heating and cooling

    Energy Technology Data Exchange (ETDEWEB)

    Swet, C.J.; Baylin, F.

    1980-07-01

    This document explores the frontier of the rapidly expanding field of thermal energy storage, investigates unresolved issues, outlines research aimed at finding solutions, and suggests avenues meriting future research. Issues related to applications include value-based ranking of storage concepts, temperature constraints, consistency of assumptions, nomenclature and taxonomy, and screening criteria for materials. Issues related to technologies include assessing seasonal storage concepts, diurnal coolness storage, selection of hot-side storage concepts for cooling-only systems, phase-change storage in building materials, freeze protection for solar water heating systems, and justification of phase-change storage for active solar space heating.

  8. Characterization of the 309 building fuel transfer pit and storage basin

    International Nuclear Information System (INIS)

    Hale, N.S.

    1998-01-01

    This document identifies radiological, chemical and physical conditions inside the Fuel Transfer Pit and Fuel Storage Basins. These spaces are located inside the Plutonium Recycle Test Reactor structure (309 Building.) The fuel handling and storage feature of the PRTR were primarily located in these spaces. The conditions were assessed as part of overall 309 Building transition

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

    International Nuclear Information System (INIS)

    MARTIN, M.K.

    2000-01-01

    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

  10. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    International Nuclear Information System (INIS)

    Burgard, K.C.

    1998-01-01

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis

  11. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-04-09

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  12. Thermal calculations for the study of the heat evacuation in the vaults building of the centralised temporary storage (ATC); Calculos termicos para el estudio de la evacuacion de calor en el Edificio de Bovedas del Almacen Temporal Centralizado (ATC)

    Energy Technology Data Exchange (ETDEWEB)

    Perez Dominguez, J. R.; Garcia Gonzalez, M.; Huelamo, E.

    2015-07-01

    This article presents the thermal analyses of the vaults building at the future Spanish Nuclear Waste Storage facility (ATC) in which spent nuclear fuel and high activity nuclear wastes are to be stored efficiency, safety and securely. the analyses have been carried out by means of computational fluid dynamics (CFD) simulation codes, for the purpose of confirming the adequate design of the storage buildings and in order to obtain the air flow rate required to guarantee that the different thermal criteria are met. The design relies on natural convection in order to remove residual heat from the nuclear waste. The simulation model allows the designer to perform sensitivity analyses to evaluate the impact of different design parameters, to optimize the heat load per fuel canister and to provide an optimal loading plan for the facility. (Author)

  13. Dry storage cells for radioactive material

    International Nuclear Information System (INIS)

    Hartley, D.J.; Paget, F.T.W.

    1982-01-01

    A facility for posting irradiated nuclear fuel from a preparation cell of a dry storage complex into storage canisters located in buckets within a clean cell comprises a telescopic tubular port member for sealably connecting the preparation cell to a canister. In operation the closure of the canister is screened against contamination and withdrawn from the canister into the preparation cell via a retractable grab prior to posting of the fuel into the canister. (author)

  14. 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

  15. Status report, canister fabrication

    International Nuclear Information System (INIS)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika; Emilsson, Goeran

    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. Integrated Building Energy Systems Design Considering Storage Technologies

    OpenAIRE

    Stadler, Michael

    2009-01-01

    The addition of storage technologies such as flow batteries, conventional batteries, and heat storage can improve the economic, as well as environmental attraction of micro-generation systems (e.g., PV or fuel cells with or without CHP) and contribute to enhanced demand response. The interactions among PV, solar thermal, and storage systems can be complex, depending on the tariff structure, load profile, etc. In order to examine the impact of storage technologies on demand response and CO2 em...

  17. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    1975-09-01

    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

  18. 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

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

    International Nuclear Information System (INIS)

    H. Marr

    1999-01-01

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

  20. K West Basin canister survey

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1998-01-01

    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

  1. Optimal control of building storage systems using both ice storage and thermal mass – Part I: Simulation environment

    International Nuclear Information System (INIS)

    Hajiah, Ali; Krarti, Moncef

    2012-01-01

    Highlights: ► A simulation environment is described to account for both passive and active thermal energy storage (TES) systems. ► Laboratory testing results have been used to validate the predictions from the simulation environment. ► Optimal control strategies for TES systems have been developed as part of the simulation environment. - Abstract: This paper presents a simulation environment that can evaluate the benefits of using simultaneously building thermal capacitance and ice storage system to reduce total operating costs including energy and demand charges while maintaining adequate occupant comfort conditions within commercial buildings. The building thermal storage is controlled through pre-cooling strategies by setting space indoor air temperatures. The ice storage system is controlled by charging the ice tank and operating the chiller during low electrical charge periods and melting the ice during on-peak periods. Optimal controls for both building thermal storage and ice storage are developed to minimize energy charges, demand charges, or combined energy and demand charges. The results obtained from the simulation environment are validated using laboratory testing for an optimal controller.

  2. 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.

  3. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    International Nuclear Information System (INIS)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo

    2016-01-01

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated

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

    International Nuclear Information System (INIS)

    Lopez, D.A.; Abbott, D.G.

    1994-02-01

    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

  5. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated.

  6. Development of the DWPF canister temporary shrink-fit seal

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-04-01

    The Defense Waste Processing Facility is being constructed at The Savannah River Plant for the containerization of high-level nuclear waste in a wasteform for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in type 304L stainless steel canisters, 2-feet in diameter x 9-feet 10-inches long, containing a flanged 6-in.-diam pipe fill-nozzle. The canisters have a minimum wall thickness of 3/8 in. Utilizing the heat from the glass filling operation, a shrink-fit seal for a plug in the end of the canister fill nozzle was developed that: will withstand the radioactive environment; will prevent the spread of contamination, and will keep moisture and water from entering the canister during storage and decontamination of the canister by wet-frit blasting to remove smearable and oxide-film fixed radioactive nuclides; is removable and can be replaced by a new oversize plug in the event the seal fails the pressure decay leakage test ( -4 atm cc/sec helium); will keep the final weld closure clean and free of nuclear contamination; will withstand being pressed into the nozzle without exposing external contamination or completely breaking the seal; is reliable; and is easily installed. The seal consists of: a removable sleeve (with a tapered bore) which is shrink-fitted into the nozzle bore during canister fabrication; and a tapered plug which is placed into the sleeved nozzle after the canister is filled with radioactive molten glass. A leak-tight shrink-fit seal is formed between the nozzle, sleeve, and plug upon temperature equilibrium. The temporarily sealed canister is transferred from the Melt cell to the Decon cell, and the surface is decontaminated. Next it is transferred to the Weld/Test cell where the temporary seal is pressed down into the nozzle, revealing a clean cavity where the canister final closure weld is made

  7. Optimal controls of building storage systems using both ice storage and thermal mass – Part II: Parametric analysis

    International Nuclear Information System (INIS)

    Hajiah, Ali; Krarti, Moncef

    2012-01-01

    Highlights: ► A detailed analysis is presented to assess the performance of thermal energy storage (TES) systems. ► Utility rates have been found to be significant in assessing the operation of TES systems. ► Optimal control strategies for TES systems can save up to 40% of total energy cost of office buildings. - Abstract: This paper presents the results of a series of parametric analysis to investigate the factors that affect the effectiveness of using simultaneously building thermal capacitance and ice storage system to reduce total operating costs (including energy and demand costs) while maintaining adequate occupant comfort conditions in buildings. The analysis is based on a validated model-based simulation environment and includes several parameters including the optimization cost function, base chiller size, and ice storage tank capacity, and weather conditions. It found that the combined use of building thermal mass and active thermal energy storage system can save up to 40% of the total energy costs when integrated optimal control are considered to operate commercial buildings.

  8. High-level waste canister envelope study: structural analysis

    International Nuclear Information System (INIS)

    1977-11-01

    The structural integrity of waste canisters, fabricated from standard weight Type 304L stainless steel pipe, was analyzed for sizes ranging from 8 to 24 in. diameter and 10 to 16 feet long under normal, abnormal, and improbable life cycle loading conditions. The canisters are assumed to be filled with vitrified high-level nuclear waste, stored temporarily at a fuel reprocessing plant, and then transported for storage in an underground salt bed or other geologic storage. In each of the three impact conditions studies, the resulting impact force is far greater than the elastic limit capacity of the material. Recommendations are made for further study

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

    International Nuclear Information System (INIS)

    FARWICK, C.C.

    1999-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  11. Identifying Critical Factors in the Cost-Effectiveness of Solar and Battery Storage in Commercial Buildings

    Energy Technology Data Exchange (ETDEWEB)

    McLaren, Joyce A. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Anderson, Katherine H. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Laws, Nicholas D. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gagnon, Pieter J. [National Renewable Energy Lab. (NREL), Golden, CO (United States); DiOrio, Nicholas A. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Li, Xiangkun [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2018-02-21

    This analysis elucidates the emerging market for distributed solar paired with battery energy storage in commercial buildings across the United States. It provides insight into the near-term and future solar and solar-plus-storage market opportunities as well as the variables that impact the expected savings from installing behind-the-meter systems.

  12. Building an organic block storage service at CERN with Ceph

    International Nuclear Information System (INIS)

    Ster, Daniel van der; Wiebalck, Arne

    2014-01-01

    Emerging storage requirements, such as the need for block storage for both OpenStack VMs and file services like AFS and NFS, have motivated the development of a generic backend storage service for CERN IT. The goals for such a service include (a) vendor neutrality, (b) horizontal scalability with commodity hardware, (c) fault tolerance at the disk, host, and network levels, and (d) support for geo-replication. Ceph is an attractive option due to its native block device layer RBD which is built upon its scalable, reliable, and performant object storage system, RADOS. It can be considered an 'organic' storage solution because of its ability to balance and heal itself while living on an ever-changing set of heterogeneous disk servers. This work will present the outcome of a petabyte-scale test deployment of Ceph by CERN IT. We will first present the architecture and configuration of our cluster, including a summary of best practices learned from the community and discovered internally. Next the results of various functionality and performance tests will be shown: the cluster has been used as a backend block storage system for AFS and NFS servers as well as a large OpenStack cluster at CERN. Finally, we will discuss the next steps and future possibilities for Ceph at CERN.

  13. Building an organic block storage service at CERN with Ceph

    Science.gov (United States)

    van der Ster, Daniel; Wiebalck, Arne

    2014-06-01

    Emerging storage requirements, such as the need for block storage for both OpenStack VMs and file services like AFS and NFS, have motivated the development of a generic backend storage service for CERN IT. The goals for such a service include (a) vendor neutrality, (b) horizontal scalability with commodity hardware, (c) fault tolerance at the disk, host, and network levels, and (d) support for geo-replication. Ceph is an attractive option due to its native block device layer RBD which is built upon its scalable, reliable, and performant object storage system, RADOS. It can be considered an "organic" storage solution because of its ability to balance and heal itself while living on an ever-changing set of heterogeneous disk servers. This work will present the outcome of a petabyte-scale test deployment of Ceph by CERN IT. We will first present the architecture and configuration of our cluster, including a summary of best practices learned from the community and discovered internally. Next the results of various functionality and performance tests will be shown: the cluster has been used as a backend block storage system for AFS and NFS servers as well as a large OpenStack cluster at CERN. Finally, we will discuss the next steps and future possibilities for Ceph at CERN.

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

    International Nuclear Information System (INIS)

    Meling, T.A.

    1995-01-01

    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. DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2001-07-30

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

  16. Storage facility for radioactive wastes

    International Nuclear Information System (INIS)

    Okada, Kyo

    1998-01-01

    Canisters containing high level radioactive wastes are sealed in overpacks in a receiving building constructed on the ground. A plurality of storage pits are formed in a layered manner vertically in multi-stages in deep underground just beneath the receiving building, for example underground of about 1000m from the ground surface. Each of the storage pits is in communication with a shaft which vertically communicates the receiving building and the storage pits, and is extended plainly in a horizontal direction from the shaft. The storage pit comprises an overpack receiving chamber, a main gallery and a plurality of galleries. A plurality of holes for burying the overpacks are formed on the bottom of the galleries in the longitudinal direction of the galleries. A plurality of overpack-positioning devices which run in the main gallery and the galleries by remote operation are disposed in the main gallery and the galleries. (I.N.)

  17. A control-oriented model for combined building climate comfort and aquifer thermal energy storage system

    NARCIS (Netherlands)

    Rostampour Samarin, Vahab; Bloemendal, J.M.; Jaxa-Rozen, M.; Keviczky, T.

    2016-01-01

    This paper presents a control-oriented model for combined building climate comfort and aquifer thermal energy storage (ATES) system. In particular, we first provide a description of building operational systems together with control framework variables. We then focus on the derivation of an

  18. Development and evaluation of a building integrated aquifer thermal storage model

    NARCIS (Netherlands)

    Bozkaya, B.; Li, R.; Labeodan, T.; Kramer, R.P.; Zeiler, W.

    2017-01-01

    An aquifer thermal energy storage (ATES) in combination with a heat pump is an excellent way to reduce the net energy usage of buildings. The use of ATES has been demonstrated to have the potential to provide a reduction of between 20 and 40% in the cooling and heating energy use of buildings. ATES

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    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

  20. Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

    International Nuclear Information System (INIS)

    Wiles, L.E.; McCann, R.A.

    1981-09-01

    This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

  1. Energy system investment model incorporating heat pumps with thermal storage in buildings and buffer tanks

    International Nuclear Information System (INIS)

    Hedegaard, Karsten; Balyk, Olexandr

    2013-01-01

    Individual compression heat pumps constitute a potentially valuable resource in supporting wind power integration due to their economic competitiveness and possibilities for flexible operation. When analysing the system benefits of flexible heat pump operation, effects on investments should be taken into account. In this study, we present a model that facilitates analysing individual heat pumps and complementing heat storages in integration with the energy system, while optimising both investments and operation. The model incorporates thermal building dynamics and covers various heat storage options: passive heat storage in the building structure via radiator heating, active heat storage in concrete floors via floor heating, and use of thermal storage tanks for space heating and hot water. It is shown that the model is well qualified for analysing possibilities and system benefits of operating heat pumps flexibly. This includes prioritising heat pump operation for hours with low marginal electricity production costs, and peak load shaving resulting in a reduced need for peak and reserve capacity investments. - Highlights: • Model optimising heat pumps and heat storages in integration with the energy system. • Optimisation of both energy system investments and operation. • Heat storage in building structure and thermal storage tanks included. • Model well qualified for analysing system benefits of flexible heat pump operation. • Covers peak load shaving and operation prioritised for low electricity prices

  2. Status of the Multipurpose Canister (MPC) Project

    International Nuclear Information System (INIS)

    Hopper, J.P.

    1996-01-01

    The multipurpose canister (MPC) project represents a cornerstone of the current U.S. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) program for handling spent nuclear fuel. The MPC and associated support equipment is being designed to accommodate the requirements for not only storage and transport but also for the specified disposal requirements of the mined geologic repository system. The phase 1 design effort for the MPC system, being performed by the Westinghouse team on behalf of TRW Environmental Safety Systems (TESS), the OCRWM management ampersand operating (M ampersand O) contractor, is on schedule for delivery of completed safety analysis reports (SARs) in April 1996

  3. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Disbrow, J.

    1994-01-01

    This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

  4. Fire hazard analysis for the fuel supply shutdown storage buildings

    International Nuclear Information System (INIS)

    REMAIZE, J.A.

    2000-01-01

    The purpose of a fire hazards analysis (FHA) is to comprehensively assess the risk from fire and other perils within individual fire areas in a DOE facility in relation to proposed fire protection so as to ascertain whether the objectives of DOE 5480.7A, Fire Protection, are met. This Fire Hazards Analysis was prepared as required by HNF-PRO-350, Fire Hazards Analysis Requirements, (Reference 7) for a portion of the 300 Area N Reactor Fuel Fabrication and Storage Facility

  5. Electricity demand and storage dispatch modeling for buildings and implications for the smartgrid

    Science.gov (United States)

    Zheng, Menglian; Meinrenken, Christoph

    2013-04-01

    As an enabler for demand response (DR), electricity storage in buildings has the potential to lower costs and carbon footprint of grid electricity while simultaneously mitigating grid strain and increasing its flexibility to integrate renewables (central or distributed). We present a stochastic model to simulate minute-by-minute electricity demand of buildings and analyze the resulting electricity costs under actual, currently available DR-enabling tariffs in New York State, namely a peak/offpeak tariff charging by consumed energy (monthly total kWh) and a time of use tariff charging by power demand (monthly peak kW). We then introduce a variety of electrical storage options (from flow batteries to flywheels) and determine how DR via temporary storage may increase the overall net present value (NPV) for consumers (comparing the reduced cost of electricity to capital and maintenance costs of the storage). We find that, under the total-energy tariff, only medium-term storage options such as batteries offer positive NPV, and only at the low end of storage costs (optimistic scenario). Under the peak-demand tariff, however, even short-term storage such as flywheels and superconducting magnetic energy offer positive NPV. Therefore, these offer significant economic incentive to enable DR without affecting the consumption habits of buildings' residents. We discuss implications for smartgrid communication and our future work on real-time price tariffs.

  6. Performance analysis of phase-change material storage unit for both heating and cooling of buildings

    Science.gov (United States)

    Waqas, Adeel; Ali, Majid; Ud Din, Zia

    2017-04-01

    Utilisation of solar energy and the night ambient (cool) temperatures are the passive ways of heating and cooling of buildings. Intermittent and time-dependent nature of these sources makes thermal energy storage vital for efficient and continuous operation of these heating and cooling techniques. Latent heat thermal energy storage by phase-change materials (PCMs) is preferred over other storage techniques due to its high-energy storage density and isothermal storage process. The current study was aimed to evaluate the performance of the air-based PCM storage unit utilising solar energy and cool ambient night temperatures for comfort heating and cooling of a building in dry-cold and dry-hot climates. The performance of the studied PCM storage unit was maximised when the melting point of the PCM was ∼29°C in summer and 21°C during winter season. The appropriate melting point was ∼27.5°C for all-the-year-round performance. At lower melting points than 27.5°C, declination in the cooling capacity of the storage unit was more profound as compared to the improvement in the heating capacity. Also, it was concluded that the melting point of the PCM that provided maximum cooling during summer season could be used for winter heating also but not vice versa.

  7. 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.

  8. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

    Furuya, Takashi; Muraoka, Susumu; Tashiro, Shingo

    1982-02-01

    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)

  9. Optimal Sizing of Energy Storage for Community Microgrids Considering Building Thermal Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Guodong [ORNL; Li, Zhi [ORNL; Starke, Michael R. [ORNL; Ollis, Ben [ORNL; Tomsovic, Kevin [University of Tennessee, Knoxville (UTK)

    2017-07-01

    This paper proposes an optimization model for the optimal sizing of energy storage in community microgrids considering the building thermal dynamics and customer comfort preference. The proposed model minimizes the annualized cost of the community microgrid, including energy storage investment, purchased energy cost, demand charge, energy storage degradation cost, voluntary load shedding cost and the cost associated with customer discomfort due to room temperature deviation. The decision variables are the power and energy capacity of invested energy storage. In particular, we assume the heating, ventilation and air-conditioning (HVAC) systems can be scheduled intelligently by the microgrid central controller while maintaining the indoor temperature in the comfort range set by customers. For this purpose, the detailed thermal dynamic characteristics of buildings have been integrated into the optimization model. Numerical simulation shows significant cost reduction by the proposed model. The impacts of various costs on the optimal solution are investigated by sensitivity analysis.

  10. Hanford Waste Vitrification Project Building limited scope risk assessment

    International Nuclear Information System (INIS)

    Braun, D.J.; Lindberg, S.E.; Reardon, M.F.; Wilson, G.P.

    1992-10-01

    A limited scope risk assessment was performed on the preliminary design of a high-level waste interim storage facility. The Canister Storage Building (CSB) facility will be built to support remediation at the US Department of Energy Hanford Site in Washington State. The CSB will be part of the support facilities for a high level Hanford Waste Vitrification Plant (HWVP). The limited scope risk assessment is based on a preliminary design which uses forced air circulation systems to move air through the building vault. The current building design calls for natural circulation to move air through the building vault

  11. Multi Canister Overpack (MCO) Topical Report [SEC 1 THRU 6

    International Nuclear Information System (INIS)

    GARVIN, L.J.

    2002-01-01

    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 facilities to permit the safe retrieval, packaging, transport, conditioning, 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 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. A common thread that was identified among the structures was the MCO. Each structure exists for the specific purpose of treating or storing the MCO and its contents. Normally, an extensive amount of MCO-related documentation would be generated for each of the facility safety analysis reports. However, the expedited schedule for removing spent fuel from the K Basins requires that the documentation effort be minimized and repetitious activities be eliminated. Therefore, this topical report has been prepared to address those aspects of the MCO that will be common to the facilities. The MCO will be included in each facility's safety documentation by reference to this topical report. By capturing the design of the MCO and its safety evaluation in a single document, repetition, inconsistency, and duplication of

  12. 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.

  13. New kinds of energy-storing building composite PCMs for thermal energy storage

    International Nuclear Information System (INIS)

    Biçer, Alper; Sarı, Ahmet

    2013-01-01

    Graphical abstract: In this work, 10 new kinds of BCPCMs were prepared by blending of liquid xylitol pentalaurate (XPL) and xylitol pentamyristate (XPM) esters into gypsum, cement, diatomite, perlite and vermiculite. DSC results showed that the melting temperatures and energy storage capacities of the prepared BCPCMs are in range of about 40–55 °C and 31–126 J/g, respectively. TG investigations and thermal cycling test showed that the BCPCMs had good thermal endurance and thermal reliability. It can be also concluded that among the prepared 10 kinds materials, especially the BCPCMs including perlite, vermiculite, diatomite were found to better candidates for thermal energy storage applications in buildings due to the fact that they have relatively high heat storage ability. Highlights: ► New kinds BCPCMs were prepared by blending of liquid XPL and XPM esters with some building materials. ► The BCPCMs had suitable melting temperatures and energy storage capacities. ► Especially, the BCPCMs including perlite, vermiculite, diatomite were found to better candidates for thermal energy storage. - Abstract: Energy storing-composite phase change materials (PCMs) are significant means of thermal energy storage in buildings. Although several building composite PCMs (BCPCMs) have been developed in recent years, the additional investigations are still required to enrich the diversity of BCPCMs for solar heating and energy conservation applications in buildings. For this purpose, the present work is focused the preparation, characterization and determination of 10 new kinds of BCPCMs. The BCPCMs were prepared by blending of liquid xylitol pentalaurate (XPL) and xylitol pentamyristate (XPM) esters with gypsum, cement, diatomite, perlite and vermiculite as supporting matrices. The scanning electron microscopy (SEM) and Fourier Transform Infrared (FT-IR) analysis showed that the ester compounds were adsorbed uniformly into the building materials due to capillary forces

  14. Cold storage with phase change material for building ventilation

    OpenAIRE

    Butala, Vincenc; Stritih, Uroš

    2015-01-01

    This paper presents an experimental and numerical analysis of building coolingusing night-time cold accumulation in phase change material (PCM), otherwise known as the "free-cooling" or "passive-cooling" principle. The phase change materials were used in ceilings and floors. The free-cooling principle is explained and some of the types of PCMs suitable for summer cooling are listed. An experiment was conducted using paraffin with a melting point of 22 °C as the PCM to store cold during the ni...

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

    International Nuclear Information System (INIS)

    Enos, David; Bryan, Charles R.

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

  17. Permitting plan for the high-level waste interim storage

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

  18. Review of thermal energy storage technologies based on PCM application in buildings

    DEFF Research Database (Denmark)

    Pomianowski, Michal Zbigniew; Heiselberg, Per; Zhang, Yinping

    2013-01-01

    Thermal energy storage systems (TES), using phase change material (PCM) in buildings, are widely investigated technologies and a fast developing research area. Therefore, there is a need for regular and consistent reviews of the published studies. This review is focused on PCM technologies...... is paid to discussion and identification of proper methods to correctly determine the thermal properties of PCM materials and their composites and as well procedures to determine their energy storage and saving potential. The purpose of the paper is to highlight promising technologies for PCM application...... developed to serve the building industry. Various PCM technologies tailored for building applications are studied with respect to technological potential to improve indoor environment, increase thermal inertia and decrease energy use for building operation. What is more, in this review special attention...

  19. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Thorpe, J.; Moore, R.S.

    1995-01-01

    The Energy Information Administration of the U.S. Department of Energy (DOE) collects data annually from commercial nuclear power reactors via the Nuclear Fuel Data survey, Form RW-859. Over the past three years, the survey has collected data on the quantities and types of nonfuel components and on the quantities and contents of canisters in storage at reactor sites. This paper focuses on the annual changes in the data, specific implications of these changes, and lessons that should be applied to future revisions of the study. The total number of canisters reported by utilities for each year from 1986 to 1993 is listed. Changes in the quantities of nonfuel components report by General Reactors from 1992 to 1993 are also provided. Comparisons of canister and nonfuel components components data from year to year and from reactor to reactor point out that survey questions on these topics have been interpreted differently by reactor personnel

  20. Heat of fusion storage systems for combined solar systems in low energy buildings

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Furbo, Simon

    2004-01-01

    Solar heating systems for combined domestic hot water and space heating has a large potential especially in low energy houses where it is possible to take full advantage of low temperature heating systems. If a building integrated heating system is used – e.g. floor heating - the supply temperature...... from solid to liquid form (Fig. 1). Keeping the temperature as low as possible is an efficient way to reduce the heat loss from the storage. Furthermore, the PCM storage might be smaller than the equivalent water storage as more energy can be stored per volume. If the PCM further has the possibility...... systems through further improvement of water based storages and in parallel to investigate the potential of using storage designs with phase change materials, PCM. The advantage of phase change materials is that large amounts of energy can be stored without temperature increase when the material is going...

  1. Development and evaluation of a building integrated aquifer thermal storage model

    DEFF Research Database (Denmark)

    Bozkaya, Basar; Li, Rongling; Labeodan, Timilehin

    2017-01-01

    An aquifer thermal energy storage (ATES) in combination with a heat pump is an excellent way to reduce the net energy usage of buildings. The use of ATES has been demonstrated to have the potential to provide a reduction of between 20 and 40% in the cooling and heating energy use of buildings. ATES...... systems are however a complex system to analyse as a number of ground conditions influence heat losses within the aquifer. ATES is also not confined from the sides and is therefore vulnerable to heat losses through conduction, advection and dispersion. The analyses of ATES system is even further...... complicated when the dynamic of a building is considered. When connected to a building, the temperature in the aquifer is influenced by the amount of heat exchange with the varying building load. Given the energy saving potentials of ATES systems in building operation, detailed understanding of the influence...

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

    International Nuclear Information System (INIS)

    C.A Kouts

    2006-01-01

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

  3. Materials for Consideration in Standardized Canister Design Activities.

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

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

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1999-01-01

    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

  5. Evaluation of existing Hanford buildings for the storage of solid wastes

    International Nuclear Information System (INIS)

    Carlson, M.C.; Hodgson, R.D.; Sabin, J.C.

    1993-05-01

    Existing storage space at the Hanford Site for solid low-level mixed waste (LLMW) will be filled up by 1997. Westinghouse Hanford Company (WHC) has initiated the project funding cycle for additional storage space to assure that new facilities are available when needed. In the course of considering the funding request, the US Department of Energy (DOE) has asked WHC to identify and review any existing Hanford Site facilities that could be modified and used as an alternative to constructing the proposed W-112 Project. This report documents the results of that review. In summary, no buildings exist at the Hanford Site that can be utilized for storage of solid LLMW on a cost-effective basis when compared to new construction. The nearest approach to an economically sensible conversion would involve upgrade of 100,000 ft 2 of space in the 2101-M Building in the 200 East Area. Here, modified storage space is estimated to cost about $106 per ft 2 while new construction will cost about $50 per ft 2 . Construction costs for the waste storage portion of the W-112 Project are comparable with W-016 Project actual costs, with escalation considered. Details of the cost evaluation for this building and for other selected candidate facilities are presented in this report. All comparisons presented address the potential decontamination and decommissioning (D ampersand D) cost avoidances realized by using existing facilities

  6. Study of Charge Build Up in UCN Storage Cell

    Science.gov (United States)

    Broering, Mark; Abney, Josh; Swank, Christopher; Filippone, Bradley; Yao, Weijun; Korsch, Wolfgang

    2017-09-01

    The neutron EDM collaboration at the Spallation Neutron Source(ORNL) is using ultra-cold neutrons in superfluid helium to improve the nEDM limit by about two orders of magnitude. These neutrons will be stored in target cells located in a strong, stable electric field. Local radiation will generate charged particles which may build up on the target cell walls reducing field strength over time. The field changes need to be kept below 1%, making it necessary to study this cell charging behavior, determine its effect on the experiment and find ways to mitigate this. In order to study this cell charging effect, a compact test setup was designed. Using this scaled down model, charged particles are generated by a 137Cs source and the electric field is monitored via the electo-optic Kerr effect. Liquid nitrogen has a much stronger response to electric fields than helium, making it an ideal candidate for first tests. Cell charging effects have been observed in liquid nitrogen. These results along with the experimental technique and progress toward a superfluid helium measurement will also be presented. This research is supported by DOE Grants: DE-FG02-99ER41101, DE-AC05-00OR22725.

  7. Peak load shifting control using different cold thermal energy storage facilities in commercial buildings: A review

    International Nuclear Information System (INIS)

    Sun, Yongjun; Wang, Shengwei; Xiao, Fu; Gao, Diance

    2013-01-01

    Highlights: • Little study reviews the load shifting control using different facilities. • This study reviews load shifting control using building thermal mass. • This study reviews load shifting control using thermal energy storage systems. • This study reviews load shifting control using phase change material. • Efforts for developing more applicable load shifting control are addressed. - Abstract: For decades, load shifting control, one of most effective peak demand management methods, has attracted increasing attentions from both researchers and engineers. Different load shifting control strategies have been developed when diverse cold thermal energy storage facilities are used in commercial buildings. The facilities include building thermal mass (BTM), thermal energy storage system (TES) and phase change material (PCM). Little study has systematically reviewed these load shifting control strategies and therefore this study presents a comprehensive review of peak load shifting control strategies using these thermal energy storage facilities in commercial buildings. The research and applications of the load shifting control strategies are presented and discussed. The further efforts needed for developing more applicable load shifting control strategies using the facilities are also addressed

  8. Operation of heat pumps for smart grid integrated buildings with thermal energy storage

    NARCIS (Netherlands)

    Finck, C.J.; Li, R.; Zeiler, W.

    2017-01-01

    A small scale office building consisting of radiant heating, a heat pump, and a water thermal energy storage tank is implemented in an optimal control framework. The optimal control aims to minimize operational electricity costs of the heat pump based on real-time power spot market prices. Optimal

  9. 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.

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-04-13

    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.

  12. 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)

    BAZINET, G.D.

    2000-01-01

    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

  13. Parametric Study on the Dynamic Heat Storage Capacity of Building Elements

    DEFF Research Database (Denmark)

    Artmann, Nikolai; Manz, H.; Heiselberg, Per

    2007-01-01

    as their interrelation. The potential of increasing thermal mass by using phase change materials (PCM) was estimated assuming increased thermal capacity. The results show a significant impact of the heat transfer coefficient on heat storage capacity, especially for thick, thermally heavy elements. The storage capacity...... of onedimensional heat conduction in a slab with convective boundary condition was applied to quantify the dynamic heat storage capacity of a particular building element. The impact of different parameters, such as slab thickness, material properties and the heat transfer coefficient was investigated, as well......In modern, extensively glazed office buildings, due to high solar and internal loads and increased comfort expectations, air conditioning systems are often used even in moderate and cold climates. Particularly in this case, passive cooling by night-time ventilation seems to offer considerable...

  14. Residential Solar-Based Seasonal Thermal Storage Systems in Cold Climates: Building Envelope and Thermal Storage

    Directory of Open Access Journals (Sweden)

    Alexandre Hugo

    2012-10-01

    Full Text Available The reduction of electricity use for heating and domestic hot water in cold climates can be achieved by: (1 reducing the heating loads through the improvement of the thermal performance of house envelopes, and (2 using solar energy through a residential solar-based thermal storage system. First, this paper presents the life cycle energy and cost analysis of a typical one-storey detached house, located in Montreal, Canada. Simulation of annual energy use is performed using the TRNSYS software. Second, several design alternatives with improved thermal resistance for walls, ceiling and windows, increased overall air tightness, and increased window-to-wall ratio of South facing windows are evaluated with respect to the life cycle energy use, life cycle emissions and life cycle cost. The solution that minimizes the energy demand is chosen as a reference house for the study of long-term thermal storage. Third, the computer simulation of a solar heating system with solar thermal collectors and long-term thermal storage capacity is presented. Finally, the life cycle cost and life cycle energy use of the solar combisystem are estimated for flat-plate solar collectors and evacuated tube solar collectors, respectively, for the economic and climatic conditions of this study.

  15. Distributed Energy Resources On-Site Optimization for Commercial Buildings with Electric and Thermal Storage Technologies

    International Nuclear Information System (INIS)

    Lacommare, Kristina S H; Stadler, Michael; Aki, Hirohisa; Firestone, Ryan; Lai, Judy; Marnay, Chris; Siddiqui, Afzal

    2008-01-01

    The addition of storage technologies such as flow batteries, conventional batteries, and heat storage can improve the economic as well as environmental attractiveness of on-site generation (e.g., PV, fuel cells, reciprocating engines or microturbines operating with or without CHP) and contribute to enhanced demand response. In order to examine the impact of storage technologies on demand response and carbon emissions, a microgrid's distributed energy resources (DER) adoption problem is formulated as a mixed-integer linear program that has the minimization of annual energy costs as its objective function. By implementing this approach in the General Algebraic Modeling System (GAMS), the problem is solved for a given test year at representative customer sites, such as schools and nursing homes, to obtain not only the level of technology investment, but also the optimal hourly operating schedules. This paper focuses on analysis of storage technologies in DER optimization on a building level, with example applications for commercial buildings. Preliminary analysis indicates that storage technologies respond effectively to time-varying electricity prices, i.e., by charging batteries during periods of low electricity prices and discharging them during peak hours. The results also indicate that storage technologies significantly alter the residual load profile, which can contribute to lower carbon emissions depending on the test site, its load profile, and its adopted DER technologies

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

    International Nuclear Information System (INIS)

    FARWICK, C.C.

    2000-01-01

    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

  17. Thermal energy storage - A review of concepts and systems for heating and cooling applications in buildings

    DEFF Research Database (Denmark)

    Pavlov, Georgi Krasimiroy; Olesen, Bjarne W.

    2012-01-01

    period required, economic viability, and operating conditions. One of the main issues impeding the utilization of the full potential of natural and renewable energy sources, e.g., solar and geothermal, for space heating and space cooling applications is the development of economically competitive......The use of thermal energy storage (TES) in buildings in combination with space heating and/or space cooling has recently received much attention. A variety of TES techniques have developed over the past decades. TES systems can provide short-term storage for peak-load shaving as well as long......-term (seasonal) storage for the introduction of natural and renewable energy sources. TES systems for heating or cooling are utilized in applications where there is a time mismatch between the demand and the most economically favorable supply of energy. The selection of a TES system mainly depends on the storage...

  18. Application of 6D Building Information Model (6D BIM) for Business-storage Building in Slovenia

    Science.gov (United States)

    Pučko, Zoran; Vincek, Dražen; Štrukelj, Andrej; Šuman, Nataša

    2017-10-01

    The aim of this paper is to present an application of 6D building information modelling (6D BIM) on a real business-storage building in Slovenia. First, features of building maintenance in general are described according to the current Slovenian legislation, and also a general principle of BIM is given. After that, step-by-step activities for modelling 6D BIM are exposed, namely from Element list for maintenance, determination of their lifetime and service measures, cost analysing and time analysing to 6D BIM modelling. The presented 6D BIM model is designed in a unique way in which cost analysis is performed as 5D BIM model with linked data to use BIM Construction Project Management Software (Vico Office), integrated with 3D BIM model, whereas time analysis as 4D BIM model is carried out as non-linked data with the help of Excel (without connection to 3D BIM model). The paper is intended to serve as a guide to the building owners to prepare 6D BIM and to provide an insight into the relevant dynamic information about intervals and costs for execution of maintenance works in the whole building lifecycle.

  19. Modeling and optimization of energy generation and storage systems for thermal conditioning of buildings targeting conceptual building design

    Energy Technology Data Exchange (ETDEWEB)

    Grahovac, Milica

    2012-11-29

    The thermal conditioning systems are responsible for almost half of the energy consump-tion by commercial buildings. In many European countries and in the USA, buildings account for around 40% of primary energy consumption and it is therefore vital to explore further ways to reduce the HVAC (Heating, Ventilation and Air Conditioning) system energy consumption. This thesis investigates the relationship between the energy genera-tion and storage systems for thermal conditioning of buildings (shorter: primary HVAC systems) and the conceptual building design. Certain building design decisions irreversibly influence a building's energy performance and, conversely, many generation and storage components impose restrictions on building design and, by their nature, cannot be introduced at a later design stage. The objective is, firstly, to develop a method to quantify this influence, in terms of primary HVAC system dimensions, its cost, emissions and energy consumption and, secondly, to enable the use of the developed method by architects during the conceptual design. In order to account for the non-stationary effects of the intermittent renewable energy sources (RES), thermal storage and for the component part load efficiencies, a time domain system simulation is required. An abstract system simulation method is proposed based on seven pre-configured primary HVAC system models, including components such as boil-ers, chillers and cooling towers, thermal storage, solar thermal collectors, and photovoltaic modules. A control strategy is developed for each of the models and their annual quasi-stationary simulation is performed. The performance profiles obtained are then used to calculate the energy consumption, carbon emissions and costs. The annuity method has been employed to calculate the cost. Optimization is used to automatically size the HVAC systems, based on their simulation performance. Its purpose is to identify the system component dimensions that provide

  20. Modeling and optimization of energy generation and storage systems for thermal conditioning of buildings targeting conceptual building design

    Energy Technology Data Exchange (ETDEWEB)

    Grahovac, Milica

    2012-11-29

    The thermal conditioning systems are responsible for almost half of the energy consump-tion by commercial buildings. In many European countries and in the USA, buildings account for around 40% of primary energy consumption and it is therefore vital to explore further ways to reduce the HVAC (Heating, Ventilation and Air Conditioning) system energy consumption. This thesis investigates the relationship between the energy genera-tion and storage systems for thermal conditioning of buildings (shorter: primary HVAC systems) and the conceptual building design. Certain building design decisions irreversibly influence a building's energy performance and, conversely, many generation and storage components impose restrictions on building design and, by their nature, cannot be introduced at a later design stage. The objective is, firstly, to develop a method to quantify this influence, in terms of primary HVAC system dimensions, its cost, emissions and energy consumption and, secondly, to enable the use of the developed method by architects during the conceptual design. In order to account for the non-stationary effects of the intermittent renewable energy sources (RES), thermal storage and for the component part load efficiencies, a time domain system simulation is required. An abstract system simulation method is proposed based on seven pre-configured primary HVAC system models, including components such as boil-ers, chillers and cooling towers, thermal storage, solar thermal collectors, and photovoltaic modules. A control strategy is developed for each of the models and their annual quasi-stationary simulation is performed. The performance profiles obtained are then used to calculate the energy consumption, carbon emissions and costs. The annuity method has been employed to calculate the cost. Optimization is used to automatically size the HVAC systems, based on their simulation performance. Its purpose is to identify the system component dimensions that provide minimal

  1. Analysis of ice cool thermal storage for a clinic building in Kuwait

    International Nuclear Information System (INIS)

    Sebzali, M.J.; Rubini, P.A.

    2006-01-01

    In Kuwait, air conditioning (AC) systems consume 61% and 40% of the peak electrical load and total electrical energy, respectively. This is due to a very high ambient temperature for the long summer period extended from April to October and the low energy cost. This paper gives an overview of the electrical peak and energy consumption in Kuwait, and it has been found that the average increase in the annual peak electrical demand and energy consumption for the year 1998-2002 was 6.2% and 6.4%, respectively. One method of reducing the peak electrical demand of AC systems during the day period is by incorporating an ice cool thermal storage (ICTS) with the AC system. A clinic building has been selected to study the effects of using an ICTS with different operation strategies such as partial (load levelling), partial (demand limiting) and full storage operations on chiller and storage sizes, reduction of peak electrical demand and energy consumption of the chiller for selected charging and discharging hours. It has been found that the full storage operation has the largest chiller and storage capacities, energy consumption and peak electrical reduction. However, partial storage (load levelling) has the smallest chiller and storage capacities and peak electrical reduction. This paper also provides a detailed comparison of using ICTS operating strategies with AC and AC systems without ICTS

  2. Energy managemant through PCM based thermal storage system for building air-conditioning: Tidel Park, Chennai

    International Nuclear Information System (INIS)

    Nallusamy, N.; Sampath, S.; Velraj, R.

    2006-01-01

    Many modern building are designed for air-conditioning and the amount of electrical energy required for providing air-conditioning can be very significant especially in the tropics. Conservation of energy is major concern to improve the overall efficiency of the system. Integration is energy storage with the conventional system gives a lot of potential for energy saving and long-term economics. Thermal energy storage systems can improve energy management and help in matching supply and demand patterns. In the present work, a detailed study has been done on the existing thermal energy storage system used in the air-conditioning system in Tidel Park, Chennai. The present study focuses on the cool energy storage system. The modes of operation and advantages of such a system for energy management are highlighted. The reason for the adoption of combined storage system and the size of the storage medium in the air-conditioning plant are analyzed. The possibility of using this concept in other cooling and heating applications, such as storage type solar water heating system, has been explored

  3. Predictive Optimal Control of Active and Passive Building Thermal Storage Inventory

    Energy Technology Data Exchange (ETDEWEB)

    Gregor P. Henze; Moncef Krarti

    2005-09-30

    Cooling of commercial buildings contributes significantly to the peak demand placed on an electrical utility grid. Time-of-use electricity rates encourage shifting of electrical loads to off-peak periods at night and weekends. Buildings can respond to these pricing signals by shifting cooling-related thermal loads either by precooling the building's massive structure or the use of active thermal energy storage systems such as ice storage. While these two thermal batteries have been engaged separately in the past, this project investigated the merits of harnessing both storage media concurrently in the context of predictive optimal control. To pursue the analysis, modeling, and simulation research of Phase 1, two separate simulation environments were developed. Based on the new dynamic building simulation program EnergyPlus, a utility rate module, two thermal energy storage models were added. Also, a sequential optimization approach to the cost minimization problem using direct search, gradient-based, and dynamic programming methods was incorporated. The objective function was the total utility bill including the cost of reheat and a time-of-use electricity rate either with or without demand charges. An alternative simulation environment based on TRNSYS and Matlab was developed to allow for comparison and cross-validation with EnergyPlus. The initial evaluation of the theoretical potential of the combined optimal control assumed perfect weather prediction and match between the building model and the actual building counterpart. The analysis showed that the combined utilization leads to cost savings that is significantly greater than either storage but less than the sum of the individual savings. The findings reveal that the cooling-related on-peak electrical demand of commercial buildings can be considerably reduced. A subsequent analysis of the impact of forecasting uncertainty in the required short-term weather forecasts determined that it takes only very

  4. PCM thermal storage design in buildings: Experimental studies and applications to solaria in cold climates

    International Nuclear Information System (INIS)

    Guarino, Francesco; Athienitis, Andreas; Cellura, Maurizio; Bastien, Diane

    2017-01-01

    Highlights: • This paper analyzes the performance of a building-integrated thermal storage system. • A wall opposing a glazed surface serves as phase change materials thermal storage. • The study is based on both experimental and simulation studies. • Heat is stored and released up to 6–8 h after solar irradiation. • Yearly heating requirements are reduced by 17% in a cold climate. - Abstract: As energy availability and demand often do not match, thermal energy storage plays a crucial role to take advantage of solar radiation in buildings: in particular, latent heat storage via phase-change material is particularly attractive due to its ability to provide high energy storage density. This paper analyzes the performance of a building-integrated thermal storage system to increase the energy performances of solaria in a cold climate. A wall opposing a highly glazed façade (south oriented) is used as thermal storage with phase change materials embedded in the wall. The study is based on both experimental and simulation studies. The concept considered is particularly suited to retrofits in a solarium since the PCM can be added as layers facing the large window on the vertical wall directly opposite. Results indicate that this PCM thermal storage system is effective during the whole year in a cold climate. The thermal storage allows solar radiation to be stored and released up to 6–8 h after solar irradiation: this has effects on both the reduction of daily temperature swings (up to 10 °C) and heating requirements (more than 17% on a yearly base). Coupling of the thermal storage system with natural ventilation is important during mid-seasons and summer to improve the PCM charge-discharge cycles and to reduce overheating. Results also show that cooling is less important than heating, reaching up to 20% of the overall annual energy requirements for the city of Montreal, Canada. Moreover, the phase change temperature range of the material used (18–24

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  6. CNAAA spent fuel complementary storage building (UFC) construction and licensing: an overview of current status

    International Nuclear Information System (INIS)

    Lima Neto, Bertino do Carmo; Pacifi, Cicero Durval

    2013-01-01

    The reprocessing of nuclear fuel assemblies could be a valuable solution in order to make available additional energy resources and also to decrease the volume of discarded materials. After the burning of nuclear fuel assemblies to produce electrical energy, these components have to be stored in the spent fuel pools of each unit, for at least 10 years, in order to decrease their residual heat. Even after this initial 10 year-period, these spent fuel assemblies still have a great amount of energy, which can be reused. Nowadays, the spent fuel materials can be reprocessed in order to produce electrical energy, or be stored to provide, in the future, an opportunity to decide how these materials will be treated. At the present moment, Brazil does not plan to reprocess these spent fuels assemblies, as performed by some other countries. Thus, Brazil intends to build a spent fuel long term intermediate storage facility to allow the chance to make a decision in the future, taking into account the available technology at that time. Considering the three CNAAA units (Angra 1, 2 and 3 of Central Nuclear Almirante Alvaro Alberto, the Brazilian nuclear power plant, located at Angra dos Reis county, Rio de Janeiro state) have a life time estimated in 60 years, and the intrinsical spent fuel pools storage capacity of these units, a Spent Fuel Complementary Storage Building - UFC has to be foreseen in order to increase the storage capacity of CNAAA. Therefore, the Spent Fuel Complementary Storage Building shall be in operation in 2018, capable to receive the first spent fuel assemblies from Angra 2 and, in the next year, from Angra 1. The same procedure will be applied for the spent fuel assemblies of Angra 3, currently in construction. The Spent Fuel Complementary Storage Building will be constructed and operated by Eletrobras Eletronuclear - the CNAAA owner - and will be located at the same site of the plant. Conceptually, the UFC will be built as a wet storage modality

  7. Doses of the staff during the spent fuel assemblies transportation and storage in Nuhmos 56V concrete system

    International Nuclear Information System (INIS)

    Atoyan, V.; Muradyan, A.

    2003-01-01

    The NUHMOS 56V concrete system provides long-term interim storage (50 years) for spent fuel assemblies, which have been out of the reactor for a sufficient period of time. It consists from horizontal storage modules. The fuel assemblies are confined in a helium atmosphere by a canister containment pressure vessel. The canister is protected and shielded by a massive reinforced concrete module. Decay heat is removed from the canister and concrete module by a passive natural draft convection ventilation system. The project of storage does not foresee the radiation monitoring inside of building and around it. But we provided and realize the radiation monitoring program around storage, it includes tree phases: - determination the zero background around the building before storage put in exploiting; - monitoring of the radioactive particles in air (additional aspiration plant); dose rate monitoring by portable dosimeters and soil monitoring during the process of the fuel storage; - constantly after the completion the fuel storage process - monitoring of the radioactive particles in air (additional aspiration plant); dose rate monitoring by portable dosimeters, and soil monitoring. Also designed the dose rate monitoring by the dosimeter RME 3 with the transfer of data by radio channel to central monitor. The canistered spent fuel assemblies are transferred from the plant's spent fuel pool to the concrete storage modules in a transfer cask. The cask is aligned with the storage module and the canister and inserted into the module by means of a hydraulic ram. The system is a totally passive installation that is designed to provide shielding and safe confinement of spent fuel for a range of postulated accident conditions and natural phenomena. (authors)

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1998-01-01

    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

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

    International Nuclear Information System (INIS)

    Pitkoff, C.C.

    1994-01-01

    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. Investigation of innovative thermochemical energy storage processes and materials for building applications

    OpenAIRE

    Aydin, Devrim

    2016-01-01

    In this study, it is aimed to develop an innovative thermochemical energy storage system through material, reactor and process based investigations for building space heating applications. The developed system could be integrated with solar thermal collectors, photovoltaic panels or heat pumps to store any excess energy in the form of heat for later use. Thereby, it is proposed to address the problem of high operational costs and CO2 emissions released by currently used fossil fuel based heat...

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

    International Nuclear Information System (INIS)

    Srinivasan, R.; Sisley, S.E.; Hopf, J.E.

    2004-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, R.; Sisley, S.E.; Hopf, J.E. [BNFL Fuel Solutions, Campbell, CA (United States)

    2004-07-01

    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.

  13. Evaluation of a molybdenum assay canister

    International Nuclear Information System (INIS)

    Yoshizumi, T.T.; Keener, S.J.

    1988-01-01

    The performance characteristics of a commercial molybdenum assay canister were evaluated. The geometrical variation of the technetium-99m (/sup 99m/Tc) activity reading was studied as a function of the elution volume for the standard vials. It was found that the /sup 99m/Tc canister activity reading was ∼ 5% lower than that of the standard method. This is due to attenuation by the canister wall. However, the effect of the geometric variation on the clinical dose preparation was found to be insignificant. The molybdenum-99 ( 99 Mo) contamination level was compared by two methods: (1) the commercial canister and (2) the standard assay kit. The 99 Mo contamination measurements with the canister indicated consistently lower readings than those with the standard 99 Mo assay kit. The authors conclude that the canister may be used in the clinical settings. However, the user must be aware of the problems and the limitations associated with this canister

  14. Experimental Study and Modeling of Ground-Source Heat Pumps with Combi-Storage in Buildings

    Directory of Open Access Journals (Sweden)

    Wessam El-Baz

    2018-05-01

    Full Text Available There is a continuous growth of heat pump installations in residential buildings in Germany. The heat pumps are not only used for space heating and domestic hot water consumption but also to offer flexibility to the grid. The high coefficient of performance and the low cost of heat storages made the heat pumps one of the optimal candidates for the power to heat applications. Thus, several questions are raised about the optimal integration and control of heat pump system with buffer storages to maximize its operation efficiency and minimize the operation costs. In this paper, an experimental investigation is performed to study the performance of a ground source heat pump (GSHP with a combi-storage under several configurations and control factors. The experiments were performed on an innovative modular testbed that is capable of emulating a ground source to provide the heat pump with different temperature levels at different times of the day. Moreover, it can emulate the different building loads such as the space heating load and the domestic hot water consumption in real-time. The data gathered from the testbed and different experimental studies were used to develop a simulation model based on Modelica that can accurately simulate the dynamics of a GSHP in a building. The model was validated based on different metrics. Energetically, the difference between the developed model and the measured values was only 3% and 4% for the heat generation and electricity consumption, respectively.

  15. Office building with an underground storage system. Operational experiences after one year; Buerogebaeude mit Erdspeicher. Betriebserfahrungen nach einem Jahr

    Energy Technology Data Exchange (ETDEWEB)

    Braun, Dorothee; Wehrli, Stefan [Basler und Hofmann AG, Zuerich (Switzerland)

    2011-07-01

    Self-sufficient heating and cooling - that was the principle of Basler paragraph Hofmann AG (Zuerich, Switzerland) and Stuecheli Architects (Zuerich, Switzerland) in the planning and constructing of a new office building in the Canton of Zuerich. For the first time an underground storage system was implemented in a commercial building. This underground storage refuels the solar energy in summer and supplies heating energy in winter. The office building was settled in in September, 2010. The pioneering project now delivers first empirical values with the underground storage system. These empirical values show: The concept comes up, but needs time.

  16. Additional guidance for including nuclear safety equivalency in the Canister Storage Building and Cold Vacuum Drying Facility final safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1997-05-20

    This document provides guidance for the production of safety analysis reports that must meet both DOE Order 5480.23 and STD 3009, and be in compliance with the DOE regulatory policy that imposes certain NRC requirements.

  17. Additional guidance for including nuclear safety equivalency in the Canister Storage Building and Cold Vacuum Drying Facility final safety analysis report

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1997-01-01

    This document provides guidance for the production of safety analysis reports that must meet both DOE Order 5480.23 and STD 3009, and be in compliance with the DOE regulatory policy that imposes certain NRC requirements

  18. Quantifying demand flexibility of power-to-heat and thermal energy storage in the control of building heating systems

    DEFF Research Database (Denmark)

    Finck, Christian; Li, Rongling; Kramer, Rick

    2018-01-01

    restricted by power-to-heat conversion such as heat pumps and thermal energy storage possibilities of a building. To quantify building demand flexibility, it is essential to capture the dynamic response of the building energy system with thermal energy storage. To identify the maximum flexibility a building......’s energy system can provide, optimal control is required. In this paper, optimal control serves to determine in detail demand flexibility of an office building equipped with heat pump, electric heater, and thermal energy storage tanks. The demand flexibility is quantified using different performance...... of TES and power-to-heat in any case of charging, discharging or idle mode. A simulation case study is performed showing that a water tank, a phase change material tank, and a thermochemical material tank integrated with building heating system can be designed to provide flexibility with optimal control....

  19. Corrective action baseline report for underground storage tank 2331-U Building 9201-1

    International Nuclear Information System (INIS)

    1994-01-01

    The purpose of this report is to provide baseline geochemical and hydrogeologic data relative to corrective action for underground storage tank (UST) 2331-U at the Building 9201-1 Site. Progress in support of the Building 9201-1 Site has included monitoring well installation and baseline groundwater sampling and analysis. This document represents the baseline report for corrective action at the Building 9201-1 site and is organized into three sections. Section 1 presents introductory information relative to the site, including the regulatory initiative, site description, and progress to date. Section 2 includes the summary of additional monitoring well installation activities and the results of baseline groundwater sampling. Section 3 presents the baseline hydrogeology and planned zone of influence for groundwater remediation

  20. A Statistical Analysis of the Economic Drivers of Battery Energy Storage in Commercial Buildings: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Long, Matthew; Simpkins, Travis; Cutler, Dylan; Anderson, Katie

    2016-11-01

    There is significant interest in using battery energy storage systems (BESS) to reduce peak demand charges, and therefore the life cycle cost of electricity, in commercial buildings. This paper explores the drivers of economic viability of BESS in commercial buildings through statistical analysis. A sample population of buildings was generated, a techno-economic optimization model was used to size and dispatch the BESS, and the resulting optimal BESS sizes were analyzed for relevant predictor variables. Explanatory regression analyses were used to demonstrate that peak demand charges are the most significant predictor of an economically viable battery, and that the shape of the load profile is the most significant predictor of the size of the battery.

  1. Horizontal modular dry irradiated fuel storage system

    Science.gov (United States)

    Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

    1988-01-01

    A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

  2. Removal Action Work Plan for 105-DR and 105-F Building Interim Safe Storage Projects and Ancillary Buildings

    International Nuclear Information System (INIS)

    Morton, M.R.

    2000-01-01

    This document contains the removal action work plan for the 105-DR and 105-F Reactor buildings and ancillary facilities. These buildings and facilities are located in the 100-D/DR and 100-F Areas of the Hanford Site, which is owned and operated by the US Department of Energy (DOE), in Benton County, Washington. The 100 Areas (including the 100-D/DR and 100-F Areas) of the Hanford Site were placed on the US Environmental Protection Agency's (EPA's) National Priorities List under the ''Comprehensive Environmental Response, Compensation, and Liability Act of 1980'' (CERCLA). The DOE has determined that hazardous substances in the 105-DR and 105-F Reactor buildings and four ancillary facilities present a potential threat to human health or the environment. The DOE has also determined that a non-time critical removal action is warranted at these facilities. Alternatives for conducting a non-time critical removal action were evaluated in the ''Engineering Evaluation/Cost Analysis for the 105-DR and 105-F Reactor Facilities and Ancillary Facilities'' (DOE-RL 1998a). The engineering evaluation/cost analysis (EE/CA) resulted in the recommendation to decontaminate and demolish the contaminated reactor buildings (except for the reactor blocks) and the ancillary facilities and to construct a safe storage enclosure (SSE) over the reactor blocks. The recommendation was approved in an action memorandum (Ecology et al. 1998) signed by the Washington State Department of Ecology (Ecology), EPA, and DOE. The DOE is the agency responsible for implementing the removal actions in the 105-D/DR and 105-F Areas. Ecology is the lead regulatory agency for facilities in the 100-D/DR Area, and EPA is the lead regulatory agency for facilities in the 100-F Area. The term ''lead regulator agency'' hereinafter, refers to these authorities. This removal action work plan supports implementation of the non-time critical removal action

  3. 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

  4. Spent fuel canister docking station

    International Nuclear Information System (INIS)

    Suikki, M.

    2006-01-01

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

  5. Dynamic Exergy Analysis for the Thermal Storage Optimization of the Building Envelope

    Directory of Open Access Journals (Sweden)

    Valentina Bonetti

    2017-01-01

    Full Text Available As a measure of energy “quality”, exergy is meaningful for comparing the potential for thermal storage. Systems containing the same amount of energy could have considerably different capabilities in matching a demand profile, and exergy measures this difference. Exergy stored in the envelope of buildings is central in sustainability because the environment could be an unlimited source of energy if its interaction with the envelope is optimised for maintaining the indoor conditions within comfort ranges. Since the occurring phenomena are highly fluctuating, a dynamic exergy analysis is required; however, dynamic exergy modelling is complex and has not hitherto been implemented in building simulation tools. Simplified energy and exergy assessments are presented for a case study in which thermal storage determines the performance of seven different wall types for utilising nocturnal ventilation as a passive cooling strategy. Hourly temperatures within the walls are obtained with the ESP-r software in free-floating operation and are used to assess the envelope exergy storage capacity. The results for the most suitable wall types were different between the exergy analysis and the more traditional energy performance indicators. The exergy method is an effective technique for selecting the construction type that results in the most favourable free-floating conditions through the analysed passive strategy.

  6. Final Hazard Classification and Auditable Safety Analysis for the 105-F Building Interim Safe Storage Project

    International Nuclear Information System (INIS)

    Rodovsky, T.J.; Bond, S.L.

    1998-07-01

    The auditable safety analysis (ASA) documents the authorization basis for the partial decommissioning and facility modifications to place the 105-F Building into interim safe storage (ISS). Placement into the ISS is consistent with the preferred alternative identified in the Record of Decision (58 FR). Modifications will reduce the potential for release and worker exposure to hazardous and radioactive materials, as well as lower surveillance and maintenance (S ampersand M) costs. This analysis includes the following: A description of the activities to be performed in the course of the 105-F Building ISS Project. An assessment of the inventory of radioactive and other hazardous materials within the 105-F Building. Identification of the hazards associated with the activities of the 105-F Building ISS Project. Identification of internally and externally initiated accident scenarios with the potential to produce significant local or offsite consequences during the 105-F Building ISS Project. Bounding evaluation of the consequences of the potentially significant accident scenarios. Hazard classification based on the bounding consequence evaluation. Associated safety function and controls, including commitments. Radiological and other employee safety and health considerations

  7. Dynamic Heat Storage and Cooling Capacity of a Concrete Deck with PCM and Thermally Activated Building System

    DEFF Research Database (Denmark)

    Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

    2012-01-01

    This paper presents a heat storage and cooling concept that utilizes a phase change material (PCM) and a thermally activated building system (TABS) implemented in a hollow core concrete deck. Numerical calculations of the dynamic heat storage capacity of the hollow core concrete deck element...... in the article highlight the potential of using TABS and PCM in a prefabricated concrete deck element....

  8. Seismic upgrading of the spent fuel storage building at Kozloduy NPP

    International Nuclear Information System (INIS)

    Alexandrov, A.; Borov, V.; Jordanov, M.; Karamanski, T.; Mihaylov, K.

    2001-01-01

    The Spent Fuel Storage Building at Kozloduy NPP site has been analysed for new review level earthquake with 0.2 g peak ground acceleration (compared to the initial design basis earthquake with 0.1 g PGA). The preliminary seismic analysis of the existing building structure using the 5% site specific response spectrum showed the need of seismic structural upgrading. Two upgrading concepts were evaluated on the basis of several factors. The main factor considered was preventing the collapse of the hall structure and the travelling cranes on the fuel storage area during and after a SSE. A three dimensional finite element model was created for the investigation of the seismic response of the existing structure and for the design of the building upgrading. The modelling of the heavy travelling crane and its sub-crane structure was one of the key points. Different configurations of the new upgrading and strengthening structures were investigated. Some interesting conclusions have been drawn from the experience in analysing and upgrading of such a complex industrial structure, comprised of elements with substantial differences in material, rigidity, construction and general behaviour. (author)

  9. 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

  10. Topical safety analysis report for the transportation of the NUHOMS reg-sign dry shielded canister

    International Nuclear Information System (INIS)

    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

  11. Energy system investment model incorporating heat pumps with thermal storage in buildings and buffer tanks

    DEFF Research Database (Denmark)

    Hedegaard, Karsten; Balyk, Olexandr

    2013-01-01

    Individual compression heat pumps constitute a potentially valuable resource in supporting wind power integration due to their economic competitiveness and possibilities for flexible operation. When analysing the system benefits of flexible heat pump operation, effects on investments should...... be taken into account. In this study, we present a model that facilitates analysing individual heat pumps and complementing heat storages in integration with the energy system, while optimising both investments and operation. The model incorporates thermal building dynamics and covers various heat storage...... of operating heat pumps flexibly. This includes prioritising heat pump operation for hours with low marginal electricity production costs, and peak load shaving resulting in a reduced need for peak and reserve capacity investments....

  12. Design premises for canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Werme, L.

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel

  13. Effect of canister size on costs of disposal of SRP high-level wastes

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-01-01

    The current plan for managing the high-level nuclear wastes at the Savannah River Plant (SRP) calls for processing them into solid forms contained in stainless steel canisters for eventual disposal in a federal geologic repository. A new SRP facility called the Defense Waste Processing Facility (DWPF) is being designed for the onsite waste processing operations. Preliminary evaluations indicate that costs of the overall disposal operation will depend significantly on the size of the canisters, which determines the number of waste forms to be processed. The objective of this study was to evaluate the effects of canister size on costs of DWPF process operations, including canister procurement, waste solidification, and interim storage, on offsite transport, and on repository costs of disposal, including provision of suitable waste packages

  14. An energy self-sufficient public building using integrated renewable sources and hydrogen storage

    International Nuclear Information System (INIS)

    Marino, C.; Nucara, A.; Pietrafesa, M.; Pudano, A.

    2013-01-01

    The control of the use of fossil fuels, major cause of greenhouse gas emissions and climate changes, in present days represents one of Governments' main challenges; particularly, a significant energy consumption is observed in buildings and might be significantly reduced through sustainable design, increased energy efficiency and use of renewable sources. At the moment, the widespread use of renewable energy in buildings is limited by its intrinsic discontinuity: consequently integration of plants with energy storage systems could represent an efficient solution to the problem. Within this frame, hydrogen has shown to be particularly fit in order to be used as an energetic carrier. In this aim, in the paper an energetic, economic and environmental analysis of two different configurations of a self-sufficient system for energy production from renewable sources in buildings is presented. In particular, in the first configuration energy production is carried out by means of photovoltaic systems, whereas in the second one a combination of photovoltaic panels and wind generators is used. In both configurations, hydrogen is used as an energy carrier, in order to store energy, and fuel cells guarantee its energetic reconversion. The analysis carried out shows that, although dimensioned as a stand-alone configuration, the system can today be realized only taking advantage from the incentivizing fares applied to grid-connected systems, that are likely to be suspended in the next future. In such case, it represents an interesting investment, with capital returns in about 15 years. As concerns economic sustainability, in fact, the analysis shows that the cost of the energy unit stored in hydrogen volumes, due to the not very high efficiency of the process, presently results greater than that of directly used one. Moreover, also the starting fund of the system proves to be very high, showing an additional cost with respect to systems lacking of energy storage equal to about 50

  15. Energy efficient hybrid nanocomposite-based cool thermal storage air conditioning system for sustainable buildings

    International Nuclear Information System (INIS)

    Parameshwaran, R.; Kalaiselvam, S.

    2013-01-01

    The quest towards energy conservative building design is increasingly popular in recent years, which has triggered greater interests in developing energy efficient systems for space cooling in buildings. In this work, energy efficient silver–titania HiTES (hybrid nanocomposites-based cool thermal energy storage) system combined with building A/C (air conditioning) system was experimentally investigated for summer and winter design conditions. HiNPCM (hybrid nanocomposite particles embedded PCM) used as the heat storage material has exhibited 7.3–58.4% of improved thermal conductivity than at its purest state. The complete freezing time for HiNPCM was reduced by 15% which was attributed to its improved thermophysical characteristics. Experimental results suggest that the effective energy redistribution capability of HiTES system has contributed for reduction in the chiller nominal cooling capacity by 46.3% and 39.6% respectively, under part load and on-peak load operating conditions. The HiTES A/C system achieved 27.3% and 32.5% of on-peak energy savings potential in summer and winter respectively compared to the conventional A/C system. For the same operating conditions, this system yield 8.3%, 12.2% and 7.2% and 10.2% of per day average and yearly energy conservation respectively. This system can be applied for year-round space conditioning application without sacrificing energy efficiency in buildings. - Highlights: • Energy storage is acquired by HiTES (hybrid nanocomposites-thermal storage) system. • Thermal conductivity of HiNPCM (hybrid nanocomposites-PCM) was improved by 58.4%. • Freezing time of HiNPCM was reduced by 15% that enabled improved energy efficiency. • Chiller nominal capacity was reduced by 46.3% and 39.6% in on-peak and part load respectively. • HiTES A/C system achieved appreciable energy savings in the range of 8.3–12.2%

  16. Criticality safety analysis of the fissile material storage arrays in the east end of building 6592

    International Nuclear Information System (INIS)

    McKeon, D.C.; Philbin, J.S.

    1981-03-01

    A criticality safety analysis of nine concrete storage holes that have been formed in the floor of the Materials Balance Area (MBA) in Building 6592 is reported. Unit cell dimensions and unit mass limits are defined for the most likely plutonium and uranium fuel types that will be stored there. Two tables of mass limits are derived. The first table is to be used for short units that can be stacked with fixed separation in the same hole. The second table will permit units greater than one foot in length providing that the appropriate linear mass density limit (in kg/ft) is not exceeded

  17. Design and building of a new experimental setup for testing hydrogen storage materials

    Energy Technology Data Exchange (ETDEWEB)

    Andreasen, Anders

    2005-09-01

    For hydrogen to become the future energy carrier a suitable way of storing hydrogen is needed, especially if hydrogen is to be used in mobile applications such as cars. To test potential hydrogen storage materials with respect to capacity, kinetics and thermodynamics the Materials Research Department has a high pressure balance. However, the drawback of this equipment is, that in order to load samples, exposure towards air is inevitable. This has prompted the design and building of a new experimental setup with a detachable reactor allowing samples to be loaded under protective atmosphere. The purpose of this report is to serve as documentation of the new setup. (au)

  18. Study of the thermal properties of selected PCMs for latent heat storage in buildings

    Science.gov (United States)

    Valentova, Katerina; Pechackova, Katerina; Prikryl, Radek; Ostry, Milan; Zmeskal, Oldrich

    2017-07-01

    The paper is focused on measurements of thermal properties of selected phase change materials (PCMs) which can be used for latent heat storage in building structures. The thermal properties were measured by the transient step-wise method and analyzed by the thermal spectroscopy. The results of three different materials (RT18HC, RT28HC, and RT35HC) and their thermal properties in solid, liquid, and phase change region were determined. They were correlated with the differential scanning calorimetry (DSC) measurement. The results will be used to determine the optimum ratio of components for the construction of drywall and plasters containing listed ingredients, respectively.

  19. Predicted peak temperature-rises around a high-level radioactive waste canister emplaced in the deep ocean bed

    International Nuclear Information System (INIS)

    Kipp, K.L.

    1978-06-01

    A simple mathematical model of heat conduction was used to evaluate the peak temperature-rise along the wall of a canister of high-level radioactive waste buried in deep ocean sediment. Three different amounts of vitrified waste, corresponding to standard Harvest, large Harvest, and AVM canisters, and three different waste loadings were studied. Peak temperature-rise was computed for the nine cases as a function of canister geometry and storage time between reprocessing and burial. Lower waste loadings or longer storage times than initially envisaged are necessary to prevent the peak temperature-rise from exceeding 200 0 C. The use of longer, thinner cylinders only modestly reduces the storage time for a given peak temperature. Effects of stacking of waste canisters and of close-packing were also studied. (author)

  20. Impact of energy storage in buildings on electricity demand side management

    International Nuclear Information System (INIS)

    Qureshi, Waqar A.; Nair, Nirmal-Kumar C.; Farid, Mohammad M.

    2011-01-01

    Research highlights: → Phase change material (PCM) application for space heating has been implemented and assessed for built environment. → Real-Time Pricing (RTP) is assessed as tool to implement Demand Side Management programs effectively. → Two buildings, with and without PCM, have been compared for space heating using RTP in functional electricity market. → PCM found to offer peak load shifting, energy conservation, and reduction in price of electricity. -- Abstract: This paper assesses impact of using phase change materials (PCM) in buildings to leverage its thermal energy storage capability. The emphasis is from an electricity demand side perspective with case studies that incorporates wholesale electricity market data of New Zealand. The results presented in this paper show that for space heating application significant advantages could be obtained using PCM built structures. These positive impacts include peak load shifting, energy conservation and reduction in peak demand for network line companies and potential reduction in electricity consumption and savings for residential customers. This paper uses a testing facility that consists of two identically designed and shaped offices built at Tamaki Campus location of the University of Auckland, New Zealand. The walls and ceilings of one office are finished with ordinary gypsum boards while the interior of the other office is finished with PCM impregnated gypsum boards. Controlled heating facility is provided in both the offices for maintaining temperature within the range of human comfort. This facility is equipped with advanced data acquisition equipment for data monitoring and archiving both locally within the offices and also remotely. Through actual observations and analysis this paper demonstrates two major impacts of DSM. First, the application of phase change material (PCM) in building environment enabling efficient thermal storage to achieve some reduction in the overall electrical energy

  1. Multi-Canister overpack inservice inspection and maintenance

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The factors to be considered in establishing inservice inspection and maintenance requirements for the Multi-Canister Overpack (MCO) include evaluating the likelihood of degradation to the MCO pressure boundary due to erosion and corrosion, reviewing commercial practice for NRC licensed spent nuclear fuel storage systems, and examining the individual MCO components for maintenance needs. Reviews of the potential for MCO erosion and corrosion conclude that neither will pose a threat to the MCO pressure boundary. Consistent with commercial practice for spent fuel storage systems, the MCO closure weld will be helium leak tested prior to placement in interim storage. Beyond the CSB facility related monitoring plans (radiological monitoring, emissions monitoring, vault cooling data, etc.), no inservice inspection or maintenance of the MCO is required during interim storage

  2. Building Energy Storage Panel Based on Paraffin/Expanded Perlite: Preparation and Thermal Performance Study.

    Science.gov (United States)

    Kong, Xiangfei; Zhong, Yuliang; Rong, Xian; Min, Chunhua; Qi, Chengying

    2016-01-25

    This study is focused on the preparation and performance of a building energy storage panel (BESP). The BESP was fabricated through a mold pressing method based on phase change material particle (PCMP), which was prepared in two steps: vacuum absorption and surface film coating. Firstly, phase change material (PCM) was incorporated into expanded perlite (EP) through a vacuum absorption method to obtain composite PCM; secondly, the composite PCM was immersed into the mixture of colloidal silica and organic acrylate, and then it was taken out and dried naturally. A series of experiments, including differential scanning calorimeter (DSC), scanning electron microscope (SEM), best matching test, and durability test, have been conducted to characterize and analyze the thermophysical property and reliability of PCMP. Additionally, the thermal performance of BESP was studied through a dynamic thermal property test. The results have showed that: (1) the surface film coating procedure can effectively solve the leakage problem of composite phase change material prepared by vacuum impregnation; (2) the optimum adsorption ratio for paraffin and EP was 52.5:47.5 in mass fraction, and the PCMP has good thermal properties, stability, and durability; and (3) in the process of dynamic thermal performance test, BESP have low temperature variation, significant temperature lagging, and large heat storage ability, which indicated the potential of BESP in the application of building energy efficiency.

  3. Thermal Energy Storage for Building Load Management: Application to Electrically Heated Floor

    Directory of Open Access Journals (Sweden)

    Hélène Thieblemont

    2016-07-01

    Full Text Available In cold climates, electrical power demand for space conditioning becomes a critical issue for utility companies during certain periods of the day. Shifting a portion or all of it to off-peak periods can help reduce peak demand and reduce stress on the electrical grid. Sensible thermal energy storage (TES systems, and particularly electrically heated floors (EHF, can store thermal energy in buildings during the off-peak periods and release it during the peak periods while maintaining occupants’ thermal comfort. However, choosing the type of storage system and/or its configuration may be difficult. In this paper, the performance of an EHF for load management is studied. First, a methodology is developed to integrate EHF in TRNSYS program in order to investigate the impact of floor assembly on the EHF performance. Then, the thermal comfort (TC of the night-running EHF is studied. Finally, indicators are defined, allowing the comparison of different EHF. Results show that an EHF is able to shift 84% of building loads to the night while maintaining acceptable TC in cold climate. Moreover, this system is able to provide savings for the customer and supplier if there is a significant difference between off-peak and peak period electricity prices.

  4. Building Energy Storage Panel Based on Paraffin/Expanded Perlite: Preparation and Thermal Performance Study

    Directory of Open Access Journals (Sweden)

    Xiangfei Kong

    2016-01-01

    Full Text Available This study is focused on the preparation and performance of a building energy storage panel (BESP. The BESP was fabricated through a mold pressing method based on phase change material particle (PCMP, which was prepared in two steps: vacuum absorption and surface film coating. Firstly, phase change material (PCM was incorporated into expanded perlite (EP through a vacuum absorption method to obtain composite PCM; secondly, the composite PCM was immersed into the mixture of colloidal silica and organic acrylate, and then it was taken out and dried naturally. A series of experiments, including differential scanning calorimeter (DSC, scanning electron microscope (SEM, best matching test, and durability test, have been conducted to characterize and analyze the thermophysical property and reliability of PCMP. Additionally, the thermal performance of BESP was studied through a dynamic thermal property test. The results have showed that: (1 the surface film coating procedure can effectively solve the leakage problem of composite phase change material prepared by vacuum impregnation; (2 the optimum adsorption ratio for paraffin and EP was 52.5:47.5 in mass fraction, and the PCMP has good thermal properties, stability, and durability; and (3 in the process of dynamic thermal performance test, BESP have low temperature variation, significant temperature lagging, and large heat storage ability, which indicated the potential of BESP in the application of building energy efficiency.

  5. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    Pitkaenen, J.

    2013-12-01

    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

  6. Safety evaluation for the inner canister closure station

    International Nuclear Information System (INIS)

    Glasscock, J.R.

    1987-01-01

    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

  7. Sacramento Municipal Utility district's interim onsite storage building for low level radioactive waste

    International Nuclear Information System (INIS)

    Gillis, E.

    1986-01-01

    In order to meet current and anticipated needs for the low level radwaste management program at the Rancho Seco Nuclear Generating Station, the Sacramento Municipal Utility District has a design and construction program underway which will provide an onsite interim storage facility that can be expanded in two and one-half year increments. The design approach utilized allows capital investment to be minimized and still provides radwaste management flexibility in anticipation of delays in resolution of the nationwide long term radwaste disposal situation. The facility provides storage and material accountability for all low level radwastes generated by the plant. Wastes are segregated by radioactivity level and are stored in two separate storage areas located within one facility. Lower activity wastes are stored in a lightly shielded structure and handled by lift trucks, while the higher activity wastes are stored in a highly shielded structure and handled remotely by manual bridge crane. The layout of the structure provides for economy of operation and minimizes personnel radiation exposure. Design philosophy and criteria, building layout and systems, estimated costs and construction schedule are discussed

  8. Thermal energy storage for building heating and cooling applications. Quarterly progress report, April--June 1976

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, H.W.; Kedl, R.J.

    1976-11-01

    This is the first in a series of quarterly progress reports covering activities at ORNL to develop thermal energy storage (TES) technology applicable to building heating and cooling. Studies to be carried out will emphasize latent heat storage in that sensible heat storage is held to be an essentially existing technology. Development of a time-dependent analytical model of a TES system charged with a phase-change material was started. A report on TES subsystems for application to solar energy sources is nearing completion. Studies into the physical chemistry of TES materials were initiated. Preliminary data were obtained on the melt-freeze cycle behavior and viscosities of sodium thiosulfate pentahydrate and a mixture of Glauber's salt and Borax; limited melt-freeze data were obtained on two paraffin waxes. A subcontract was signed with Monsanto Research Corporation for studies on form-stable crystalline polymer pellets for TES; subcontracts are being negotiated with four other organizations (Clemson University, Dow Chemical Company, Franklin Institute, and Suntek Research Associates). Review of 10 of 13 unsolicited proposals received was completed by the end of June 1976.

  9. Canister arrangement for storing radioactive waste

    Science.gov (United States)

    Lorenzo, D.K.; Van Cleve, J.E. Jr.

    1980-04-23

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

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

    International Nuclear Information System (INIS)

    Michael B. Heiser; Clark B. Millet

    2005-01-01

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

  11. Effect of Canister Movement on Water Turbidity

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    2000-01-01

    Requirements for evaluating the adherence characteristics of sludge on the fuel stored in the K East Basin and the effect of canister movement on basin water turbidity are documented in Briggs (1996). The results of the sludge adherence testing have been documented (Bergmann 1996). This report documents the results of the canister movement tests. The purpose of the canister movement tests was to characterize water turbidity under controlled canister movements (Briggs 1996). The tests were designed to evaluate methods for minimizing the plumes and controlling water turbidity during fuel movements leading to multi-canister overpack (MCO) loading. It was expected that the test data would provide qualitative visual information for use in the design of the fuel retrieval and water treatment systems. Video recordings of the tests were to be the only information collected

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

    International Nuclear Information System (INIS)

    Sande, W.E.; Bjorklund, W.J.; Brooks, N.A.

    1977-04-01

    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

  13. Stakeholder involvement in the evaluation of a multipurpose canister system

    International Nuclear Information System (INIS)

    Williams, J.R.; Kane, D.; Smith, T.B. Jr.

    1994-01-01

    The U.S. Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), began evaluating a multipurpose canister (MPC) concept in October of 1992. This followed recommendations by the Nuclear Waste Technical Review Board (NWTRB) and the U.S. Nuclear Regulatory Commission (NRC) that DOE develop a nuclear waste management system that achieves system integration, standardization, and reduced fuel-handling operations. Industry organizations such as Edison Electric Institute (EEI) and Electric Power Research Institute (EPRI) had conducted earlier studies that concluded advantages to the nuclear waste management system may be offered by such a concept. The MPC concept involves a metal canister which would contain multiple spent nuclear fuel assemblies. The canister would be sealed at the nuclear power plant and would not be reopened. The MPC would then be placed inside separate casks or overpacks for storage, transportation, and disposal. An important factor in DOE's evaluation of the MPC concept was the involvement of external parties. This paper describes that involvement process for the OCRWM's MPC implementation program. External parties who have an interest or stake in the program are referred to as stakeholders

  14. Decontamination of stainless steel canisters that contain high-level waste

    International Nuclear Information System (INIS)

    Bray, L.A.

    1987-01-01

    At the West Valley Nuclear Services Company (WVNSC) in West Valley, New York, high-level radioactive waste (HLW) will be vitrified into a borosilicate glass form and poured into large, stainless steel canisters. During the filling process, volatile fission products, principally 137 Cs, condense on the exterior of the canisters. The smearable contamination that remains on the canisters after they are filled and partially cooled must be removed from the canisters' exterior surfaces prior to their storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. A simple and effective method was developed for decontamination of HLW canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. The process employs a reduction-oxidation system [Ce(III)/Ce(IV)] in nitric acid solution to chemically mill the surface of stainless steel, similar to the electropolishing process, but without the need for an applied electrical current. Contaminated canisters are simply immersed in the solution at controlled temperature and Ce(IV) concentration levels

  15. 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.

  16. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Beesley. J.F.

    2005-01-01

    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

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

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Faletti, D.W.; Budden, M.J.

    1988-03-01

    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

  18. Thermodynamic analyses and assessments of various thermal energy storage systems for buildings

    International Nuclear Information System (INIS)

    Caliskan, Hakan; Dincer, Ibrahim; Hepbasli, Arif

    2012-01-01

    Highlights: ► Proposing a novel latent (PCM), thermochemical and sensible (aquifer) TES combination for building heating. ► Performing comprehensive environmental, energy, exergy and sustainability analyses. ► Investigating the effect of varying dead state temperatures on the TESs. - Abstract: In this study, energetic, exergetic, environmental and sustainability analyses and their assessments are carried out for latent, thermochemical and sensible thermal energy storage (TES) systems for phase change material (PCM) supported building applications under varying environment (surrounding) temperatures. The present system consists of a floor heating system, System-I, System-II and System-III. The floor heating system stays at the building floor supported with a floor heating unit and pump. The System-I includes a latent TES system and a fan. The latent TES system is comprised of a PCM supported building envelope, in which from outside to inside; glass, transparent insulation material, PCM, air channel and insulation material are placed, respectively. Furthermore, System-II mainly has a solar-thermochemical TES while there are an aquifer TES and a heat pump in System-III. Among the TESs, the hot and cold wells of the aquifer TES have maximum exergetic efficiency values of 88.782% and 69.607% at 8 °C dead state temperature, respectively. According to the energy efficiency aspects of TESs, the discharging processes of the latent TES and the hot well of the aquifer TES possess the minimum and maximum values of 5.782% and 94.118% at 8 °C dead state temperature, respectively. Also, the fan used with the latent TES is the most environmentally-benign system component among the devices. Furthermore, the most sustainable TES is found for the aquifer TES while the worst sustainable system is the latent TES.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  20. Thermal energy storage characteristics of bentonite-based composite PCMs with enhanced thermal conductivity as novel thermal storage building materials

    International Nuclear Information System (INIS)

    Sarı, Ahmet

    2016-01-01

    Graphical abstract: In this work, novel bentonite-based and form-stable composite phase change materials (Bb-FSPCMs) were produced for LHTES in buildings by impregnation of CA, PEG600, DD and HD with bentonite clay. The microstructures of the compatibility of the Bb-FSPCMs were by using SEM and FT-IR techniques. The DSC results indicated that the produced Bb-FSPCMs composites had suitable phase change temperature of 4–30 °C and good latent heat capacity between 38 and 74 J/g. The TG results demonstrated that all of the fabricated Bb-FSPCMs had good thermal resistance. The Bb-FSPCMs maintained their LHTES properties even after 1000 heating–cooling cycling. The total heating times of the prepared Bb-FSPCMs were reduced noticeably due to their enhanced thermal conductivity after EG (5 wt%) addition. - Highlights: • Bb-FSPCMs were produced by impregnation of CA, PEG600, DD and HD with bentonite. • DSC analysis indicated that Bb-FSPCMs had melting temperature in range of 4–30 °C. • DSC analysis also showed that Bb-FSPCMs had latent heat between 38 and 74 J/g. • The TG analysis demonstrated that Bb-FSPCMs had good thermal resistance. • Thermal conductivity of Bb-FSPCMs were enhanced noticeably with EG (5 wt%) addition. - Abstract: In this work, for latent heat thermal energy storage (LHTES) applications in buildings, bentonite-based form-stable composite phase change materials (Bb-FSPCMs) were produced by impregnation of capric acid (CA), polyethylene glycol (PEG600), dodecanol (DD) and heptadecane (HD) into bentonite clay. The morphological characterization results obtained by scanning electron microscopy (SEM) showed that the bentonite acted as good structural barrier for the organic PCMs homogenously dispersed onto its surface and interlayers. The chemical investigations made by using fourier transform infrared (FT-IR) technique revealed that the attractions between the components of the composites was physical in nature and thus the PCMs were hold

  1. 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.

  2. Dynamic thermal behavior of building using phase change materials for latent heat storage

    Directory of Open Access Journals (Sweden)

    Selka Ghouti

    2015-01-01

    Full Text Available This study presents a two-dimensional model with a real size home composed of two-storey (ground and first floor spaces separated by a slab, enveloped by a wall with rectangular section containing phase change material (PCM in order to minimize energy consumption in the buildings. The main objective of the PCM-wall system is to decrease the temperature change from outdoor space before it reaches the indoor space during the daytime. The numerical approach uses effective heat capacity Ceff model with realistic outdoor climatic conditions of Tlemcen city, Algeria. The numerical results showed that by using PCM in wall as energy storage components may reduce the room temperature by about 6 to 7°C of temperature depending on the floor level (first floor spaces or ground floor spaces.

  3. 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.

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

    International Nuclear Information System (INIS)

    Geuther, J.; Conrad, E.A.; Rhoadarmer, D.

    2009-01-01

    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

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    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

  6. Spent nuclear fuel storage - Basic concept

    International Nuclear Information System (INIS)

    Krempel, Ascanio; Santos, Cicero D. Pacifici dos; Sato, Heitor Hitoshi; Magalhaes, Leonardo de

    2009-01-01

    According to the procedures adopted in others countries in the world, the spent nuclear fuel elements burned to produce electrical energy in the Brazilian Nuclear Power Plant of Angra do Reis, Central Nuclear Almirante Alvaro Alberto - CNAAA will be stored for a long time. Such procedure will allow the next generation to decide how they will handle those materials. In the future, the reprocessing of the nuclear fuel assemblies could be a good solution in order to have additional energy resource and also to decrease the volume of discarded materials. This decision will be done in the future according to the new studies and investigations that are being studied around the world. The present proposal to handle the nuclear spent fuel is to storage it for a long period of time, under institutional control. Therefore, the aim of this paper is to introduce a proposal of a basic concept of spent fuel storage, which involves the construction of a new storage building at site, in order to increase the present storage capacity of spent fuel assemblies in CNAAA installation; the concept of the spent fuel transportation casks that will transfer the spent fuel assemblies from the power plants to the Spent Fuel Complementary Storage Building and later on from this building to the Long Term Intermediate Storage of Spent Fuel; the concept of the spent fuel canister and finally the basic concept of the spent fuel long term storage. (author)

  7. Fabrication and properties of microencapsulated-paraffin/gypsum-matrix building materials for thermal energy storage

    International Nuclear Information System (INIS)

    Su Junfeng; Wang Xinyu; Wang Shengbao; Zhao Yunhui; Huang Zhen

    2012-01-01

    Graphical abstract: DSC curves of microPCMs/gypsum composite samples before and after a thermal cycling treatment. Highlights: ► Microcapsules containing paraffin was fabricated by in-situ polymerization. ► Methanol-modified melamine–formaldehyde (MMF) was used as shell material. ► MicroPCMs/gypsum-matrix building materials were applied for solar energy storage. ► The structure and thermal conductivity of composites had been investigated. - Abstract: Microencapsulated phase change materials (microPCMs) have been widely applied in solid matrix as thermal-storage or temperature-controlling functional composites. The aim of this work was to prepare and investigate the properties of microPCMs/gypsum-matrix building materials for thermal energy storage. MicroPCMs contain paraffin was fabricated by in situ polymerization using methanol-modified melamine–formaldehyde (MMF) as shell material. A series of microPCMs samples were prepared under emulsion stirring rates in range of 1000–3000 r min −1 with core/shell weight ratios of 3/1, 2/1, 1/1, 1/2 and 1/3, respectively. The shell of microPCMs was smooth and compact with global shape, its thickness was not greatly affected by the core/shell ratio and emulsion stirring rate. DSC tests showed that the shell of microPCMs did not influence the phase change behavior of pure paraffin. It was found from TGA analysis that microPCMs samples containing paraffin lost their weight at the temperature of nearly 250 °C, which indicated that the PCM had been protected by shell. More shell material in microPCMs could enhance the thermal stability and provide higher compact condition for core material. After a 100-times thermal cycling treatment, the microPCMs contain paraffin also nearly did not change the phase change behaviors of PCM. With the increasing of weight contents of microPCMs in gypsum board, the thermal conductivity (λ) values of composites had decreased. The simulation of temperature tests proved that the

  8. Modeling and simulation to determine the potential energy savings by implementing cold thermal energy storage system in office buildings

    International Nuclear Information System (INIS)

    Rismanchi, B.; Saidur, R.; Masjuki, H.H.; Mahlia, T.M.I.

    2013-01-01

    Highlights: • Simulating the CTES system behavior based on Malaysian climate. • Almost 65% of power is used for cooling for cooling the office buildings, every day. • The baseline shows an acceptable match with real data from the fieldwork. • Overall, the energy used for full load storage is much than the conventional system. • The load levelling storage strategy has 3.7% lower energy demand. - Abstract: In Malaysia, air conditioning (AC) systems are considered as the major energy consumers in office buildings with almost 57% share. During the past decade, cold thermal energy storage (CTES) systems have been widely used for their significant economic benefits. However, there were always doubts about their energy saving possibilities. The main objective of the present work is to develop a computer model to determine the potential energy savings of implementing CTES systems in Malaysia. A case study building has been selected to determine the energy consumption pattern of an office building. In the first step the building baseline model was developed and validated with the recorded data from the fieldwork. Once the simulation results reach an acceptable accuracy, different CTES system configuration was added to the model to predict their energy consumption pattern. It was found that the overall energy used by the full load storage strategy is considerably more than the conventional system. However, by applying the load leveling storage strategy, and considering its benefits to reduce the air handling unit size and reducing the pumping power, the overall energy usage was almost 4% lower than the non-storage system. Although utilizing CTES systems cannot reduce the total energy consumption considerably, but it has several outstanding benefits such as cost saving, bringing balance in the grid system, reducing the overall fuel consumption in the power plants and consequently reducing to total carbon footprint

  9. Analytical Evaluation of Preliminary Drop Tests Performed to Develop a Robust Design for the Standardized DOE Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Ware, A.G.; Morton, D.K.; Smith, N.L.; Snow, S.D.; Rahl, T.E.

    1999-01-01

    The Department of Energy (DOE) has developed a design concept for a set of standard canisters for the handling, interim storage, transportation, and disposal in the national repository, of DOE spent nuclear fuel (SNF). The standardized DOE SNF canister has to be capable of handling virtually all of the DOE SNF in a variety of potential storage and transportation systems. It must also be acceptable to the repository, based on current and anticipated future requirements. This expected usage mandates a robust design. The canister design has four unique geometries, with lengths of approximately 10 feet or 15 feet, and an outside nominal diameter of 18 inches or 24 inches. The canister has been developed to withstand a drop from 30 feet onto a rigid (flat) surface, sustaining only minor damage - but no rupture - to the pressure (containment) boundary. The majority of the end drop-induced damage is confined to the skirt and lifting/stiffening ring components, which can be removed if de sired after an accidental drop. A canister, with its skirt and stiffening ring removed after an accidental drop, can continue to be used in service with appropriate operational steps being taken. Features of the design concept have been proven through drop testing and finite element analyses of smaller test specimens. Finite element analyses also validated the canister design for drops onto a rigid (flat) surface for a variety of canister orientations at impact, from vertical to 45 degrees off vertical. Actual 30-foot drop testing has also been performed to verify the final design, though limited to just two full-scale test canister drops. In each case, the analytical models accurately predicted the canister response

  10. Application of cold thermal energy storage (CTES) for building demand management in hot climates

    International Nuclear Information System (INIS)

    Comodi, Gabriele; Carducci, Francesco; Nagarajan, Balamurugan; Romagnoli, Alessandro

    2016-01-01

    Highlights: • A new index, Savings per energy unit, is defined to assess the effectiveness of CTES. • CTES systems were used to perform demand management strategies, removing partial load operations and shaving peak loads. • CTES was used to perform price arbitrage, exploiting the difference between peak and off peak electricity rates in Singapore. • Results showed that it is possible to enhance the efficiency of the whole system, achieving both energy and economic savings. • Depending on the sizing scenario, the pay back periods ranged from a minimum of 8.9 years to a maximum of 16 years. - Abstract: This paper investigates the feasibility of Cold Thermal Energy Storage (CTES) for building demand management applications in hot climate characterized by a cooling season lasting all year long. An existing office building, located in Singapore, serves as case study. The CTES is coupled to the existing cooling systems in order to address the opportunity of improving overall energy efficiency and to perform price arbitrage, exploiting the spread between peak and off-peak energy tariffs. Six different sizes for the CTES are analyzed, addressing different percentage of the daily cooling energy demand. A new index, Savings per energy unit, is defined to assess the effectiveness of CTES. Results indicate that it is possible to enhance the efficiency of the whole cooling system, achieving both energy and economic savings. The payback periods of the different solutions range from a minimum of 8.9 years to a maximum of 16 years. All these aspects make CTES applications a viable option. However, a large amount of space in direct proximity to the building is necessary and, especially in largely urban environment, this is not always available.

  11. Canister transfer into repository in shaft alternative

    International Nuclear Information System (INIS)

    Raiko, H.; Kukkola, T.; Autio, J.

    2005-09-01

    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.)

  12. Utilizing thermal building mass for storage in district heating systems: Combined building level simulations and system level optimization

    DEFF Research Database (Denmark)

    Dominkovic, D. F.; Gianniou, P.; Münster, M.

    2018-01-01

    on the energy supply of district heating. Results showed that longer preheating time increased the possible duration of cut-off events. System optimization showed that the thermal mass for storage was used as intra-day storage. Flexible load accounted for 5.5%–7.7% of the total district heating demand...

  13. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

    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

  14. Design premises for canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Werme, L

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel 43 refs, 4 figs, 6 tabs

  15. Criticality safety studies for the storage of waste from nuclear fuel service in Intercell Storage Wells 2 and 3 of Building 3019

    International Nuclear Information System (INIS)

    Primm, R.T. III; Hopper, C.M.; Smolen, G.R.

    1992-11-01

    This report provides computational evaluation results demonstrating that mixed oxide waste can be safely stored in Intercell Storage Wells 2 and 3 of Building 3019 at the Oak Ridge National Laboratory. Existing, verified computational techniques are validated with applicable critical experiments and tolerance limits for safety analyses are derived. Multiplication factors for normal and credible abnormal configurations are calculated and found to be far subcritical when compared to derived safety limits

  16. Grain boundary corrosion of copper canister material

    International Nuclear Information System (INIS)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J.

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

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

  18. A review on energy conservation in building applications with thermal storage by latent heat using phase change materials

    International Nuclear Information System (INIS)

    Khudhair, Amar M.; Farid, Mohammed M.

    2004-01-01

    Energy storage in the walls, ceiling and floor of buildings may be enhanced by encapsulating suitable phase change materials (PCMs) within these surfaces to capture solar energy directly and increase human comfort by decreasing the frequency of internal air temperature swings and maintaining the temperature closer to the desired temperature for a longer period of time. This paper summarizes the investigation and analysis of thermal energy storage systems incorporating PCMs for use in building applications. Researches on thermal storage in which the PCM is encapsulated in concrete, gypsum wallboard, ceiling and floor have been ongoing for some time and are discussed. The problems associated with the application of PCMs with regard to the selection of materials and the methods used to contain them are also discussed

  19. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

    Baich, M.A.

    2000-01-01

    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. Fatty acid esters-based composite phase change materials for thermal energy storage in buildings

    International Nuclear Information System (INIS)

    Sarı, Ahmet; Karaipekli, Ali

    2012-01-01

    In this study, fatty acid esters-based composite phase change materials (PCMs) for thermal energy storage were prepared by blending erythritol tetrapalmitate (ETP) and erythritol tetrastearate (ETS) with diatomite and expanded perlite (EP). The maximum incorporation percentage for ETP and ETS into diatomite and EP was found to be 57 wt% and 62 wt%, respectively without melted PCM seepage from the composites. The morphologies and compatibilities of the composite PCMs were structurally characterized using scanning electron microscope (SEM) and Fourier transformation infrared (FT–IR) analysis techniques. Thermal energy storage properties of the composite PCMs were determined by differential scanning calorimetry (DSC) analysis. The DSC analyses results indicated that the composite PCMs were good candidates for building applications in terms of their large latent heat values and suitable phase change temperatures. The thermal cycling test including 1000 melting and freezing cycling showed that composite PCMs had good thermal reliability and chemical stability. TG analysis revealed that the composite PCMs had good thermal durability above their working temperature ranges. Moreover, in order to improve the thermal conductivity of the composite PCMs, the expanded graphite (EG) was added to them at different mass fractions (2%, 5%, and 10%). The best results were obtained for the composite PCMs including 5wt% EG content in terms of the increase in thermal conductivity values and the decrease amount in latent heat capacity. The improvement in thermal conductivity values of ETP/Diatomite, ETS/Diatomite, ETP/EP and ETS/EP were found to be about 68%, 57%, 73% and 75%, respectively. Highlights: ► Fatty acid esters-based composite PCMs were prepared by blending ETP and ETS with diatomite and expanded perlite. ► The composite PCMs were characterized by using SEM, FT–IR, DSC and TG analysis methods. ► The DSC results indicated that the composites PCMs had good thermal

  1. 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.

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

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah; Wati

    2007-01-01

    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/mm 2 . (author)

  3. 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

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

    International Nuclear Information System (INIS)

    Lydmark, Sara; Pedersen, Karsten

    2011-03-01

    '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 10 0 -10 2 viable sulphate-reducing and acetogenic bacteria and 10 2 -10 4 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

  5. Processes, Techniques, and Successes in Welding the Dry Shielded Canisters of the TMI-2 Reactor Core Debris

    International Nuclear Information System (INIS)

    Zirker, L.R.; Rankin, R.A.; Ferrell, L.J.

    2002-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is operated by Bechtel-BWXT Idaho LLC (BBWI), which recently completed a very successful $100 million Three-Mile Island-2 (TMI-2) program for the Department of Energy (DOE). This complex and challenging program used an integrated multidisciplinary team approach that loaded, welded, and transported an unprecedented 25 dry shielded canisters (DSC) in seven months, and did so ahead of schedule. The program moved over 340 canisters of TMI-2 core debris that had been in wet storage into a dry storage facility at the INEEL. The main thrust of this paper is relating the innovations, techniques, approaches, and lessons learned associated to welding of the DSC's. This paper shows the synergism of elements to meet program success and shares these lessons learned that will facilitate success with welding of dry shielded canisters in other DOE complex dry storage programs

  6. Preliminary design for spent fuel canister handling systems in a canister transfer and installation vehicle

    International Nuclear Information System (INIS)

    Wendelin, T.; Suikki, M.

    2008-12-01

    The report presents a spent fuel canister transfer and installation vehicle. The vehicle is used for carrying the fuel canister into a disposal tunnel and installing it into a deposition hole. The report outlines basic requirements and a design for canister handling equipment used in a canister transfer and installation vehicle, a description regarding the operation and maintenance of the equipment, as well as a cost estimate. Specific vehicles will be manufactured for all canister types in order to minimize the height of the disposal tunnels. This report is only focused on a transfer and installation vehicle for OL1-2 fuel canisters. Detailed designing and selection of final components have not yet been carried out. The report also describes the vehicle's requirements for the structures of a repository system, as well as actions in possible malfunction or fault situations. The spent fuel canister is brought from an encapsulation plant by a canister lift down to the repository level. The fuel canister is driven from the canister lift by an automated guided vehicle onto a canister hoist at a canister loading station. The canister transfer and installation vehicle is waiting for the canister with its radiation shield in an upright position above the canister hoist. The hoist carries the canister upward until the vehicle's own lifting means grab hold of the canister and raise it up into the vehicle's radiation shield. This is followed by turning the radiation shield to a transport position and by closing it in a radiation-proof manner against a rear radiation shield. The vehicle is driven along the central tunnel into the disposal tunnel and parked on top of the deposition hole. The vehicle's radiation shield is turned to the upright position and the canister is lowered with the vehicle's hydraulic winches into a bentonite-lined deposition hole. The radiation shield is turned back to the transport position and the vehicle can be driven out of the disposal tunnel

  7. Evaluation of salt particle collection device for preventing SCC on canister - Effect on particle collection rate by electric field

    International Nuclear Information System (INIS)

    Takeda, H.; Saegusa, T.

    2013-01-01

    Now, in Japan, while metal casks are used for spent nuclear fuel storage, a practical use of concrete casks is under review because of its cost effectiveness and procurement easiness. In reviewing the practical use, stress corrosion cracking (SCC) of a canister container in the concrete cask becomes an issue and is needed to be resolved soon. A natural ventilation system is generally adopted for the storage facilities, especially in Japan where facilities are built near coasts so that the cooling air includes sea salt particles. Therefore, the occurrence of SCC is concerned when the sea salt particles adhere to welded parts of the canisters. In this study, we proposed a salt particle collection device with low pressure loss which does not interfere with the air flow into the building or the concrete casks. The device is composed of a stack of 10 parallel stainless steel plates, the air is free to circulate in the space between them. Pressure loss tests in a laboratory and salt particle collection tests in the field have been performed. It has been clarified that the pressure loss of the device is one-thirtieth to one-twentieth of that of a commercial filter and 40% of the particles in the air could be collected and the device would not influence the heat removal performance. Moreover, we evaluated the effect of electric field on the particle collection under supposing the particle charge. In the case of electric field over 10 3 kV/m the particle collection rate could be improved dramatically

  8. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

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

  9. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    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

  10. 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.)

  11. 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.

  12. Remote controlled mover for disposal canister transfer

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

    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.)

  13. Functional and operational requirements document : building 1012, Battery and Energy Storage Device Test Facility, Sandia National Laboratories, New Mexico.

    Energy Technology Data Exchange (ETDEWEB)

    Johns, William H. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2013-11-01

    This report provides an overview of information, prior studies, and analyses relevant to the development of functional and operational requirements for electrochemical testing of batteries and energy storage devices carried out by Sandia Organization 2546, Advanced Power Sources R&D. Electrochemical operations for this group are scheduled to transition from Sandia Building 894 to a new Building located in Sandia TA-II referred to as Building 1012. This report also provides background on select design considerations and identifies the Safety Goals, Stakeholder Objectives, and Design Objectives required by the Sandia Design Team to develop the Performance Criteria necessary to the design of Building 1012. This document recognizes the Architecture-Engineering (A-E) Team as the primary design entity. Where safety considerations are identified, suggestions are provided to provide context for the corresponding operational requirement(s).

  14. Decommissioning of four small nuclear waste storage buildings and an evaporation plant

    International Nuclear Information System (INIS)

    Hedvall, R.H.; Ellmark, C.; Stocker, P.

    2008-01-01

    A small-scale decommissioning concept was applied with staff from an earlier project wish strong knowledge of radiation protection, minimized radiation doses and environmental pollution. The project was therefore initiated with less than 10 people involved using standard hand held equipment. The aim of the decommissioning project was to set free as much material as possible, i.e. remove waste from the regulatory control regime and also free the remaining structures and buildings for conventional demolition and subsequent reuse of the property. Complete decommissioning will be concluded at the end of 2008 when all waste is taken case of. This is the fourth in a series of important decommissioning projects in Studsvik since the 1980s. Some of the conclusions are: 1) Obtain a group with well-known personnel that have been working together before for the entire project For a project larger than this, project management assistant would have made follow-up more efficient. Experts in instrumentation and statistics are also important. Also important is knowledge about practical decisions that would make the project more efficient in terms of time. Interviews and historical facts are important when choosing which nuclides are of most interest for measurements (but be critic). 2) Be sure all authoritative requirements are followed, like setting up a work environment plan at the entrance to the site and placing a fence around the work site. 3) Check all individual radiation exposures before project start and do whole body measurements both before and after the project. Urine samples should be taken if alpha contamination is a risk. 4) Calculate for unwanted and 'not what you expected' situations in the time schedule. 5) Be aware of contaminations and radiation sources outside the actual area. They might have to be moved. 6) Calculate and order bins and containers for waste storage well in advance. Stay informed of the updated amount of waste and keep it in locked storage. 7

  15. Heat of Fusion Storage with High Solar Fraction for Solar Low Energy Buildings

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Furbo, Simon

    2006-01-01

    to achieve 100% coverage of space heating and domestic hot water in a low energy house in a Danish climate with a solar heating system with 36 m² flat plate solar collector and approximately 10 m³ storage with sodium acetate. A traditional water storage solution aiming at 100% coverage will require a storage...... of the storage to cool down below the melting point without solidification preserving the heat of fusion energy. If the supercooled storage reaches the surrounding temperature no heat loss will take place until the supercooled salt is activated. The investigation shows that this concept makes it possible...

  16. Revised corrective action plan for underground storage tank 2331-U at the Building 9201-1 Site

    International Nuclear Information System (INIS)

    Bohrman, D.E.; Ingram, E.M.

    1993-09-01

    This document represents the Corrective Action Plan for underground storage tank (UST) 2331-U, previously located at Building 9201-1, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee. Tank 2331-U, a 560-gallon UST, was removed on December 14, 1988. This document presents a comprehensive summary of all environmental assessment investigations conducted at the Building 9201-1 Site and the corrective action measures proposed for remediation of subsurface petroleum product contamination identified at the site. This document is written in accordance with the regulatory requirements of the Tennessee Department of Environment and Conservation (TDEC) Rule 1200-1-15-.06(7)

  17. Friction welded closures of waste canisters

    International Nuclear Information System (INIS)

    Klein, R.F.

    1987-01-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it into a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive-seal weld, the properties and thickness of which must be at least equal to those of the canister material. All studies and tests performed in the work discussed in this paper have the inertia friction welding concept to be highly feasible in this application. This paper describes the decision to investigate the inertia friction welding process, the inertia friction welding process itself, and a proposed equipment design concept. This system would provide a positive, reliable, inspectable, and full-thickness seal weld while utilizing easily maintainable equipment. This high-quality weld can be achieved even in highly contaminated hot cell

  18. Corrosion resistance of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban

    2007-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  19. Corrosion resistance of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban [Corrosion and Metals Research Institute, Sto ckholm (Sweden)

    2007-03-15

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  20. System Specification for Immobilized High-Level Waste Interim Storage

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    2000-01-01

    This specification establishes the system-level functional, performance, design, interface, and test requirements for Phase 1 of the IHLW Interim Storage System, located at the Hanford Site in Washington State. The IHLW canisters will be produced at the Hanford Site by a Selected DOE contractor. Subsequent to storage the canisters will be shipped to a federal geologic repository

  1. Dry storage cell for radioactive material

    International Nuclear Information System (INIS)

    Bradley, N.

    1982-01-01

    In a dry storage cell for irradiated nuclear fuel or other highly active waste, cooling air flow is by natural draught in heat exchange with fuel containing canisters housed in channels. To inhibit corrosion by ensuring that the temperature of the air flowing over the canisters does not fall below the dew point when heat generation by decay has fallen, a fraction of the heat energy transferred to the cooling air is recirculated to the air upstream of the canisters. Recirculation of heat energy is effected by recirculation of a fraction of the hot air from downstream of the canisters. (author)

  2. Storage, transportation and disposal system for used nuclear fuel assemblies

    Science.gov (United States)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  3. Removal Action Workplan for the 105-DR and 105-F Building Interim Safe Storage Projects and Ancillary Buildings

    International Nuclear Information System (INIS)

    Rodovsky, T. J.

    1999-01-01

    This document is the removal action workplan (RAW) for the 105-DR and 105-F Reactor Buildings and ancillary facilities. These buildings and facilities are located in the 100-D/DR and 100-F Areas of the Hanford Site in Benton County, Washington, which is owned and operated by the U.S. Department of Energy (DOE). The 100 Areas (including 100-D/DR and 100-F Areas) of the Hanford Site were placed on the U.S. Environmental Protection Agency's (EPA) National Priorities List under the Comprehensive Environmental Response, Compensation,and Liability Act of 1980 (CERCLA). DOE has determined that hazardous substances in the 105-DR and 105-F Reactor Buildings and four ancillary facilities present a potential threat to human health or the environment DOE has also determined that a non-time critical removal action is warranted at these facilities

  4. Building arrangement and site layout design guides for on site low level radioactive waste storage facilities

    International Nuclear Information System (INIS)

    McMullen, J.W.; Feehan, M.J.

    1986-01-01

    Many papers have been written by AE's and utilities describing their onsite storage facilities, why they are needed, NRC regulations, and disposal site requirements. This paper discusses a typical storage facility and address the design considerations and operational aspects that are generally overlooked when designing and siting a low level radioactive waste storage facility. Some topics to be addressed are: 1. Container flexibility; 2. Modular expansion capabilities; 3. DOT regulations; 4. Meterological requirements; 5. OSHA; 6. Fire protection; 7. Floods; 8. ALARA

  5. Copper canisters for nuclear high level waste disposal. Corrosion aspects

    International Nuclear Information System (INIS)

    Werme, L.; Sellin, P.; Kjellbert, N.

    1992-10-01

    A corrosion analysis of a thick-walled copper canister for spent fuel disposal is discussed. The analysis has shown that there are no rapid mechanisms that may lead to canister failure, indicating an anticipated corrosion service life of several millions years. If further analysis of the copper canister is considered, it should be concentrated on identifying and evaluating processes other than corrosion, which may have a potential for leading to canister failure. (au)

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

    International Nuclear Information System (INIS)

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

    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

  7. Thermal studies of the canister staging pit in a hypothetical Yucca Mountain canister handling facility using computational fluid dynamics

    International Nuclear Information System (INIS)

    Soltani, Mehdi; Barringer, Chris; Bues, Timothy T. de

    2007-01-01

    The proposed Yucca Mountain nuclear waste storage site will contain facilities for preparing the radioactive waste canisters for burial. A previous facility design considered was the Canister Handling Facility Staging Pit. This design is no longer used, but its thermal evaluation is typical of such facilities. Structural concrete can be adversely affected by the heat from radioactive decay. Consequently, facilities must have heating ventilation and air conditioning (HVAC) systems for cooling. Concrete temperatures are a function of conductive, convective and radiative heat transfer. The prediction of concrete temperatures under such complex conditions can only be adequately handled by computational fluid dynamics (CFD). The objective of the CFD analysis was to predict concrete temperatures under normal and off-normal conditions. Normal operation assumed steady state conditions with constant HVAC flow and temperatures. However, off-normal operation was an unsteady scenario which assumed a total HVAC failure for a period of 30 days. This scenario was particularly complex in that the concrete temperatures would gradually rise, and air flows would be buoyancy driven. The CFD analysis concluded that concrete wall temperatures would be at or below the maximum temperature limits in both the normal and off-normal scenarios. While this analysis was specific to a facility design that is no longer used, it demonstrates that such facilities are reasonably expected to have satisfactory thermal performance. (author)

  8. Building

    OpenAIRE

    Seavy, Ryan

    2014-01-01

    Building for concrete is temporary. The building of wood and steel stands against the concrete to give form and then gives way, leaving a trace of its existence behind. Concrete is not a building material. One does not build with concrete. One builds for concrete. MARCH

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

    International Nuclear Information System (INIS)

    Werme, L.

    1990-09-01

    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.)

  10. CSER 79-028, Addendum 2: Security bar addition to pedestal storage racks in Room 3 in 2736-Z Building

    International Nuclear Information System (INIS)

    Miller, E.M.

    1994-01-01

    The Plutonium Finishing Plant (PFP) is installing security bars on plutonium storage racks in Room 3 in 2736-Z Building to meet International Atomic Energy Agency (IAEA) material control requirements. Figures show the existing arrangement and design of the security bars. The security bars are to be fabricated of aluminum or carbon steel. The detailed fabrication sketches are reproduced in Appendix C. The security bars are to be installed close to the chains of plutonium so a determination of their effect on criticality safety needs to be made. The addition of security bars to the storage array of 2.5 kg plutonium buttons in Room 3 can effect reactivity by reflecting neutrons back into the plutonium in the storage cans, by absorbing neutrons, and by moderating neutrons between stored plutonium buttons. The small amount of metal added by the storage bars in comparison to the amount of concrete in the walls and aluminum in the shelf monitors already in place would not significantly increase the k eff of the storage array. Several computer calculations in previous analyses show that the security bars will have a negligible affect on reactivity

  11. Plutonium Immobilization Project - Robotic canister loading

    International Nuclear Information System (INIS)

    Hamilton, R.L.

    2000-01-01

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF)

  12. Techniques for freeing deposited canisters. Final report

    International Nuclear Information System (INIS)

    Kalbantner, P.; Sjoeblom, R.

    2000-06-01

    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

  13. Decontamination of high-level waste canisters

    International Nuclear Information System (INIS)

    Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

    1980-12-01

    This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces

  14. Investigation of heat of fusion storage for solar low energy buildings

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Furbo, Simon

    2005-01-01

    This paper describes a theoretical investigation by means of TRNSYS simulations of a partly heat loss free phase change material (PCM) storage solution for solar heating systems. The partly heat loss free storage is obtained by controlled used of super cooling in a mixture of sodium acetate...

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

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    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

  16. Investigation on Solar Heating System with Building-Integrated Heat Storage

    DEFF Research Database (Denmark)

    Heller, Alfred

    1996-01-01

    Traditional solar heating systems cover between 5 and 10% of the heat demand fordomestic hot water and comfort heating. By applying storage capacity this share can beincreased much. The Danish producer of solar heating systems, Aidt-Miljø, markets such a system including storage of dry sand heated...... by PP-pipe heat exchanger. Heat demand is reduced due to direct solar heating and due to storage. The storage affects the heat demand passively due to higher temperatures. Hence heat loss is reduced and passive heating is optioned. In theory, by running the system flow backwards, active heating can...... solar collector area of the system, was achieved. Active heating from the sand storage was not observed. The pay-back time for the system can be estimated to be similar to solar heated domestic hot water systems in general. A number of minor improvements on the system could be pointed out....

  17. Advanced storage concepts for solar thermal systems in low energy buildings. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Furbo, S.; Andersen, Elsa; Schultz, Joergen M.

    2006-04-07

    The aim of Task 32 is to develop new and advanced heat storage systems which are economic and technical suitable as long-term heat storage systems for solar heating plants with a high degree of coverage. The project is international and Denmark's participation has focused on Subtask A, C, and D. In Subtask A Denmark has contributed to a status report about heat storage systems. In Subtask C Denmark has focused on liquid thermal storage tanks based on NaCH{sub 3}COO?3H{sub 2}O with a melting point of 58 deg. C. Theoretical and experimental tests have been conducted in order to establish optimum conditions for storage design. In Subtask D theoretical and experimental tests of optimum designs for advanced water tanks for solar heating plants for combined space heating and domestic hot water have been conducted. (BA)

  18. Heat of Fusion Storage with High Solar Fraction for Solar Low Energy Buildings

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Furbo, Simon

    The paper presents the results of a theoretical investigation of use of phase change materials (PCM’s) with active use of super cooling as a measure for obtaining partly heat loss free seasonal storages for solar combi-systems with 100% coverage of the energy demand of both space heating and dome......The paper presents the results of a theoretical investigation of use of phase change materials (PCM’s) with active use of super cooling as a measure for obtaining partly heat loss free seasonal storages for solar combi-systems with 100% coverage of the energy demand of both space heating...... and domestic hot water. The work is part of the IEA Solar Heating & Cooling Programme Task 32 “Advanced Storage Concepts for Solar Buildings”. The investigations are based on a newly developed TRNSYS type for simulation of a PCM-storage with controlled super-cooling. The super-cooling makes it possible to let...... storage parts already melted to cool down to surrounding temperature without solidification in which state that part of the storage will be heat loss free but still will hold the latent heat in form of the heat of fusion. At the time of energy demand the solidification of the super-cooled storage part...

  19. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1996-12-01

    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. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    1996-12-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

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

    International Nuclear Information System (INIS)

    Werme, L.

    1991-12-01

    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

  2. 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.

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

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    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.)

  4. Annual Collection and Storage of Solar Energy for the Heating of Buildings, Report No. 3. Semi-Annual Progress Report, August 1977 - January 1978.

    Science.gov (United States)

    Beard, J. Taylor; And Others

    This report is part of a series from the Department of Energy on the use of solar energy in heating buildings. Described here is a new system for year around collection and storage of solar energy. This system has been operated at the University of Virginia for over a year. Composed of an underground hot water storage system and solar collection,…

  5. Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 and 2

    International Nuclear Information System (INIS)

    CARRELL, R.D.

    2002-01-01

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

  6. Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    CARRELL, R D

    2002-07-16

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft{sup 2} and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available.

  7. The utilization of the storage of thermal energy in buildings. Underground heat storages - thermic simulation and profitability; Termisen energian varastoinnin hyvaeksikaeyttoemahdollisuudet rakennusten laemmityksessae ja jaeaehdytyksessae. Maanalaiset varastot - laempoetekninen simulointi ja taloudellinen kannattavuus

    Energy Technology Data Exchange (ETDEWEB)

    Suokas, M.; Heinonen, J.; Karola, A.; Laine, T.; Siren, K.

    1998-12-31

    Interest in different sources of free energy has significantly increased due to the possibility to decrease the consumption of fossil fuels and nuclear power. This can be reached, for example, with waste heat recovery and by utilising natural heat and cool energy sources. The main problem is that the supply and use of energy do not encounter and this causes a need for thermal energy storage. The earlier heat storage systems have utilised compressor heat pumps because the temperature levels of heat storages are not high enough for the ordinary heating and cooling systems. The disadvantage is the complexity of these systems which leads to increasing building costs. Therefore, this study deals with systems of low temperature levels used mainly for cooling purposes. The aim was to find out their usability, savings and profitability. The function and energy consumption of systems were simulated with models of buildings, soil heat storage and climate. The soil model simulates heat dynamic behaviour of the masses of soil. With the climate model it was possible to simulate transient heat losses of the storage and building. It was also possible to simulate various climatic conditions by changing input data of the climate model. In the simulated systems the emphasis is on the production of cooling energy by utilising the low temperature of the ground. The systems consist of heat storage and building. The cooling energy will be charged in winter to the storage when the heat energy charged in summer will be transferred to the supply air of ventilating unit. After the energy simulations the investment and usage costs of this kind of systems were compared with costs of ordinary compressor cooling systems. The buildings studied were an imaginary LVIS 2000 office building and the Messukeskus in Helsinki which is a large hall built for exhibitions. The types of soil were wet clay and granite. The LVIS 2000 office building needs a rock heat storage with capacity of 8 000-30 000 m

  8. Energetic and Exergy Efficiency of a Heat Storage Unit for Building Heating

    International Nuclear Information System (INIS)

    Hazami, Mejdi; Kooli, Sami; Lazaar, Meriem; Farhat, Abdelhamid; Belghith, Ali

    2009-01-01

    This paper deals with a numerical and experimental investigation of a daily solar storage system conceived and built in Laboratoire de Maitrise des Technologies de l Energie (LMTE, Borj Cedria). This system consists mainly of the storage unit connected to a solar collector unit. The storage unit consists of a wooden case with dimension of 5 m 3 (5 m x 1m x 1m) filed with fin sand. Inside the wooden case was buried a network of a polypropylene capillary heat exchanger with an aperture area equal to 5 m 2 . The heat collection unit consisted of 5 m 2 of south-facing solar collector mounted at a 37 degree tilt angle. In order to evaluate the system efficiency during the charging period (during the day) and discharging period (during the night) an energy and exergy analyses were applied. Outdoor experiments were also carried out under varied environmental conditions for several consecutive days. Results showed that during the charging period, the average daily rates of thermal energy and exergy stored in the heat storage unit were 400 and 2.6 W, respectively. It was found that the net energy and exergy efficiencies in the charging period were 32 pour cent and 22 pour cent, respectively. During the discharging period, the average daily rates of the thermal energy and exergy recovered from the heat storage unit were 2 kW and 2.5 kW, respectively. The recovered heat from the heat storage unit was used for the air-heating of a tested room (4 m x 3 m x 3 m). The results showed that 30 pour cent of the total heating requirement of the tested room was obtained from the heat storage system during the whole night in cold seasons

  9. A model predictive framework of Ground Source Heat Pump coupled with Aquifer Thermal Energy Storage System in heating and cooling equipment of a building

    NARCIS (Netherlands)

    Rostampour Samarin, V.; Bloemendal, J.M.; Keviczky, T.

    2017-01-01

    This paper presents a complete model of a building heating and cooling equipment and a ground source heat pump (GSHP) coupled with an aquifer thermal energy storage (ATES) system. This model contains detailed
    mathematical representations of building thermal dynamics, ATES system dynamics, heat

  10. Economic assessment of energy storage for load shifting in Positive Energy Building

    DEFF Research Database (Denmark)

    Dumont, Olivier; Carmo, Carolina; Georges, Emeline

    2016-01-01

    Net Zero Energy Buildings (NZEB) and Positive Energy Buildings (PEB) are gaining more and more interest. In this paper, the impact of the integration of a battery in a positive energy building is assessed in order to increase its self-consumption of electricity. Parametric studies are carried out......-in tariff and a 5 kWh battery. Finally, simple correlations (based on the feed-in tariff, the annual electrical consumption and production) to predict the optimal size of battery and the lowest payback period are proposed.......Net Zero Energy Buildings (NZEB) and Positive Energy Buildings (PEB) are gaining more and more interest. In this paper, the impact of the integration of a battery in a positive energy building is assessed in order to increase its self-consumption of electricity. Parametric studies are carried out...... by varying the building envelope characteristics, the power supply system, the climate, the lightning and appliances profiles, the roof tilt, the battery size and the electricity tariffs, leading to 3200 cases. The analysis is performed on an annual basis in terms of self-consumption rate, shifted energy...

  11. Economic assessment of electric energy storage for load shifting in positive energy building

    DEFF Research Database (Denmark)

    Dumont, Olivier; Do Carmo, Carolina Madeira Ramos; Georges, Emeline

    2017-01-01

    Net zero energy buildings and positive energy buildings are gaining more and more interest. This paper evaluates the impact of the integration of a battery in a positive energy building used to increase its self-consumption of electricity. Parametric studies are carried out by varying the buildin...... and a 3.7 kWh battery. Finally, simple correlations (based on the feed-in tariff, the annual electrical consumption and production) to predict the optimal size of battery and the lowest payback period are proposed.......Net zero energy buildings and positive energy buildings are gaining more and more interest. This paper evaluates the impact of the integration of a battery in a positive energy building used to increase its self-consumption of electricity. Parametric studies are carried out by varying the building...... envelope characteristics, the power supply system, the climate, the lighting and appliances profiles, the roof tilt angle, the battery size and the electricity tariffs, leading to 3200 cases. The analysis is performed on an annual basis in terms of self-consumption and self-production rate and payback...

  12. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-06-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  13. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    J.S. Tang

    2000-01-01

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Design Section. The objective of this calculation is to determine the source terms that include radionuclide inventory, decay heat, and radiation sources due to gamma rays and neutrons for the high-level radioactive waste (HLW) from the, West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HS), and Idaho National Engineering and Environmental Laboratory (INEEL). This calculation also determines the source terms of the canister containing the SRS HLW glass and immobilized plutonium. The scope of this calculation is limited to source terms for a time period out to one million years. The results of this calculation may be used to carry out performance assessment of the potential repository and to evaluate radiation environments surrounding the waste packages (WPs). This calculation was performed in accordance with the Development Plan ''Source Terms for HLW Glass Canisters'' (Ref. 7.24)

  14. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant liquid, high-level radioactive waste into a solid form, such as borosilicate glass. To prevent the spread of radioactivity, the outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated by-products, which are difficult to immobilize by vitrification

  15. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-01-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  16. Criticality design evaluation of the White Sands reactor building storage vault

    International Nuclear Information System (INIS)

    Philbin, J.S.; Nelson, W.E.

    1979-03-01

    This report describes the conceptual design and criticality evaluation of a storage vault for components of the fast pulse reactor at White Sands Missile Range. Criticality calculations were performed with the KENO-IV Monte Carlo code for various storage configurations in order to investigate the coupling between the portable reactor and storage arrays of spare reactor rings or other fissile components of similar mass. Abnormal conditions corresponding to pseudo--random arrays of the fuel components, as well as a number of flooded configurations, were also evaluated to assess criticality potential for highly unlikely situations. In a normal, dry configuration, the neutron self-multiplication factor, k/sub eff/, of the fully loaded 3 x 8 planar array plus the reactor is less than 0.87. A completely flooded vault was found to produce self-multiplication factors in excess of 1.2

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-02

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

  18. Supervisory Control of Loads and Energy Storage in Next-Generation Zero Energy Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kung, Feitau [National Renewable Energy Lab. (NREL), Golden, CO (United States); Frank, Stephen [National Renewable Energy Lab. (NREL), Golden, CO (United States); Scheib, Jennifer [National Renewable Energy Lab. (NREL), Golden, CO (United States); Bernal Heredia, Willy [National Renewable Energy Lab. (NREL), Golden, CO (United States); Pless, Shanti [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-09-01

    A zero energy building (ZEB)—also known as a net zero energy or zero net energy building—is a building that exports as much renewable energy as the total energy it imports from other sources on an annual basis (DOE 2015). Large-scale and commercially viable ZEBs are now in the marketplace, and they are expected to become a larger share of the commercial building footprint as government and private sector policies continue to promote the development of buildings that produce more on-site energy than they use. However, the load profiles of ZEBs are currently perceived by electric utilities to be unfavorable and unpredictable. As shown in Figure ES-1, ZEB load profiles can have abrupt changes in magnitude, at times switching rapidly between exporting and importing electricity. This is a challenge for utilities, which are responsible for constantly balancing electricity supply and demand across the grid. Addressing these concerns will require new strategies and tools.

  19. Evaluation of a fast power demand response strategy using active and passive building cold storages for smart grid applications

    International Nuclear Information System (INIS)

    Cui, Borui; Wang, Shengwei; Yan, Chengchu; Xue, Xue

    2015-01-01

    Highlights: • A fast power demand response strategy is developed for smart grid applications. • The developed strategy can provide immediate and stepped power demand reduction. • The demand reduction and building indoor temperature can be predicted accurately. • The demand reduction during the DR event is stable. - Abstract: Smart grid is considered as a promising solution in improving the power reliability and sustainability where demand response is one important ingredient. Demand response (DR) is a set of demand-side activities to reduce or shift electricity use to improve the electric grid efficiency and reliability. This paper presents the investigations on the power demand alternation potential for buildings involving both active and passive cold storages to support the demand response of buildings connected to smart grids. A control strategy is developed to provide immediate and stepped power demand reduction through shutting chiller(s) down when requested. The primary control objective of the developed control strategy is to restrain the building indoor temperature rise as to maintain indoor thermal comfort within certain level during the DR event. The chiller power reduction is also controlled under certain power reduction set-point. The results show that stepped and significant power reduction can be achieved through shutting chiller(s) down when requested. The power demand reduction and indoor temperature during the DR event can be also predicted accurately. The power demand reduction is stable which is predictable for the system operators

  20. Dry storage systems with free convection air cooling

    International Nuclear Information System (INIS)

    Kioes, S.R.

    1980-01-01

    Several design principles to remove heat from the spent fuel by free air convection are illustrated and described. The key safety considerations were felt to be: loss of coolant is impossible as the passive system uses air as a coolant; overheating is precluded because as the temperatures of the containers rises the coolant flow rate increases; mass of the storage building provides a large heat sink and therefore a rapid temperature rise is impossible; and lack of any active external support requirements makes the cooling process less likely to equipment or operator failures. An example of this type of storage already exists. The German HTGR is operated with spherical graphite fuel elements which are stored in canister and in storage cells. The concept is a double cooling system with free convection inside the cells and heat exchange via two side walls of the cell to the ambient air in the cooling ducts. Technical description of the TN 1300 cask is also presented

  1. 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)

    McClelland, B.C.; Moore, T.D.

    2006-01-01

    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)

  2. 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

  3. Characterization of plutonium-containing materials and storage canisters

    International Nuclear Information System (INIS)

    Mason, R.E.

    1997-01-01

    Throughout the weapons complex, plutonium materials are stored in various containers. Some plutonium has been stored for 20 yr or more. The physical and chemical properties of the plutonium material and the containers that hold it are often not well characterized. The U.S. Department of Energy (DOE) 3013 standard sets criteria to which stored material must conform. The 3013 standard regulates materials that hold 50% or greater plutonium, and the other 50% is not specified and is usually unknown. The Materials Identification and Surveillance project is tasked to characterize representative materials and begin to characterize the other 50% and to show that materials can be brought into 3013 criteria conformance through thermal treatments

  4. 'Eco-house 99' - Full-scale demonstration of solar walls with building integrated heat storages

    Energy Technology Data Exchange (ETDEWEB)

    Hummelshoej, R.M.; Rahbek, J.E. [COWI Consulting Engineers and Planners AS (Denmark)

    2000-07-01

    A critical issue for solar systems in northern latitudes is the economic profitability. It is often said that the techniques for solar utilisation are expensive and unprofitable. This is, however, not always the case. A new project with 59 low energy terrace houses was carried out in Kolding, Denmark. The houses are designed as ecological buildings with emphasis on total economy based on low operation and maintenance costs, energy conservation and passive/hybrid solar utilisation. Besides direct solar gain through windows, each house has a solar wall of 6-8.5 m{sup 2} on the south facade. The solar walls are used both for heating of ventilation air and for space heating. The solar walls deliver heat to the dwellings during the heating season. To optimise the energy utilisation from the solar walls, the energy is stored internally in building integrated heat storages. Two different new types of prefabricated heat storages are built into the houses. One is an internal concrete wall with embedded ventilation pipes, and the other is a hollow concrete element with integrated stone bed. The heat storages are mainly designed to store solar energy from the day to the evening and the night. Because the solar walls and the heat storages have been a part of the design process from the start, the additional expenses are as low as 30-140 Euro/m{sup 2} solar wall compared with the alternative facade. This is far less than what it costs to add a solar wall on an existing building. Measurements over one year show that the yield of the solar walls is in the range of 115-125 kWh/m{sup 2}/year as expected. With the actual financing, the annual payment of the additional expenses for the solar systems is between 1-6 Euro/m{sup 2} solar wall, while the annual savings are about 5 Euro/year/m{sup 2} (with an energy price of 0.042 Euro/kWh). Dependent on which alternative facade construction the solar wall system is compared with, the profit of the system is in the range of 1 to +4 Euro

  5. Stress corrosion cracking of copper canisters

    International Nuclear Information System (INIS)

    King, Fraser; Newman, Roger

    2010-12-01

    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

  6. 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

  7. Development of spent fuel dry storage technology

    International Nuclear Information System (INIS)

    Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

    2000-01-01

    The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

  8. <90 day storage training plan for the 103-B, 1701-BA, AND 1714-C buildings

    International Nuclear Information System (INIS)

    Snider, P.M.

    1996-10-01

    The 103-B, 1701-BA, and 1714C < 90 Day Storage Area stores characteristic wastes generated in the demolition of the 103-B, 1701-BA, and 1714-C Complex. Wastes (lead-based painted components) are packaged and stored in vendor shipment containers. This is the Environmental Restoration Contractor team training plan for the 103-B, 1701-BA, and 1714-C subgrade demolition < 90 Day Storage of Hazardous Waste. This document is intended to meet the requirements of Washington Administrative Code 173-303-330 and the Hanford Dangerous Waste Permit. Training unrelated to compliance with WAC 173-303-330 is not addressed in this training plan. WAC 173-303-330(1)(d)(2, 5, 6) requires that personnel be familiarized, where applicable, with waste feed cut-off systems, response to ground-water contamination incidents, and shutdown of operations. These are not applicable to 103-B, 1701-BA, and 1714-C Subgrade Demolition < 90 Day Storage, and therefore are not covered in this training plan

  9. Fracturing of simulated high-level waste glass in canisters

    International Nuclear Information System (INIS)

    Peters, R.D.; Slate, S.C.

    1981-09-01

    Waste-glass castings generated from engineering-scale developmental processes at the Pacific Northwest Laboratory are generally found to have significant levels of cracks. The causes and extent of fracturing in full-scale canisters of waste glass as a result of cooling and accidental impact are discussed. Although the effects of cracking on waste-form performance in a repository are not well understood, cracks in waste forms can potentially increase leaching surface area. If cracks are minimized or absent in the waste-glass canisters, the potential for radionuclide release from the canister package can be reduced. Additional work on the effects of cracks on leaching of glass is needed. In addition to investigating the extent of fracturing of glass in waste-glass canisters, methods to reduce cracking by controlling cooling conditions were explored. Overall, the study shows that the extent of glass cracking in full-scale, passively-cooled, continuous melting-produced canisters is strongly dependent on the cooling rate. This observation agrees with results of previously reported Pacific Northwest Laboratory experiments on bench-scale annealed canisters. Thus, the cause of cracking is principally bulk thermal stresses. Fracture damage resulting from shearing at the glass/metal interface also contributes to cracking, more so in stainless steel canisters than in carbon steel canisters. This effect can be reduced or eliminated with a graphite coating applied to the inside of the canister. Thermal fracturing can be controlled by using a fixed amount of insulation for filling and cooling of canisters. In order to maintain production rates, a small amount of additional facility space is needed to accomodate slow-cooling canisters. Alternatively, faster cooling can be achieved using the multi-staged approach. Additional development is needed before this approach can be used on full-scale (60-cm) canisters

  10. Can-in-canister cold demonstration in DWPF (U)

    International Nuclear Information System (INIS)

    Kuehn, N.H.

    1996-07-01

    The Department of Energy Fissile Materials Disposition Program is evaluating a number of options for disposition of weapons-usable plutonium surplus to national defense needs. One of the immobilization options is the Can-In-Canister approach. In this option small cans of a plutonium glass, which contains a neutron absorber, are placed on a support structure in a large Savannah River Site Defense Waste Processing Facility (DWPF) canister. The top is then welded onto the canister. This canister is filled with High Level Waste (HLW) glass at the DWPF. The HLW glass provides the radiation source for proliferation resistance. These canisters are to be placed in a Federal Repository. To provide information on the technical feasibility of this option prior to the Record of Decision on plutonium disposition, the Department of Energy Fissile Materials Disposition Program funded a demonstration in the DWPF. This demonstration was conducted before the start of radioactive operations. Two test canisters containing cans of surrogate (non- radioactive) plutonium glass were successfully filled with simulated HLW glass at the DWPF using standard pouring procedures. One canister had twenty cans of surrogate plutonium glass. The other had eight cans of surrogate plutonium glass. After the canisters were filled, the contents of the canisters were examined to provide data on the effect of the rack and cans on the filling of the DWPF canister, the effect of the pour on the surrogate plutonium glass and the effect of the rack and cans on the simulated HLW glass. There was no deformation of the support racks during the pour. The simulated HLW glass filled all the regions around the rack and cans and the regions between the cans and the wall of the canister. This report discusses the design of the racks and cans, the modification of the DWPF canisters to accommodate the rack and cans, the conditions during the pours and the results of the post pour analysis

  11. Clathrate Hydrates for Thermal Energy Storage in Buildings: Overview of Proper Hydrate-Forming Compounds

    Directory of Open Access Journals (Sweden)

    Beatrice Castellani

    2014-09-01

    Full Text Available Increasing energy costs are at the origin of the great progress in the field of phase change materials (PCMs. The present work aims at studying the application of clathrate hydrates as PCMs in buildings. Clathrate hydrates are crystalline structures in which guest molecules are enclosed in the crystal lattice of water molecules. Clathrate hydrates can form also at ambient pressure and present a high latent heat, and for this reason, they are good candidates for being used as PCMs. The parameter that makes a PCM suitable to be used in buildings is, first of all, a melting temperature at about 25 °C. The paper provides an overview of groups of clathrate hydrates, whose physical and chemical characteristics could meet the requirements needed for their application in buildings. Simulations with a dynamic building simulation tool are carried out to evaluate the performance of clathrate hydrates in enhancing thermal comfort through the moderation of summer temperature swings and, therefore, in reducing energy consumption. Simulations suggest that clathrate hydrates have a potential in terms of improvement of indoor thermal comfort and a reduction of energy consumption for cooling. Cooling effects of 0.5 °C and reduced overheating hours of up to 1.1% are predicted.

  12. Energy flexibility of residential buildings using short term heat storage in the thermal mass

    DEFF Research Database (Denmark)

    Dreau, Jerome Le; Heiselberg, Per Kvols

    2016-01-01

    Highlights •Two residential buildings (80's and passive house) with two emitters (radiator, UH). •Different modulations of the set-point (upward/downward, duration, starting time). •Large differences between the 80s and the passive house, influence of the emitter. •Evaluation of the flexibility...

  13. PFPF canister counter for foreign plutonium (PCAS-3) hardware operations and procedures manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Baca, J.; Kroncke, K.E.; Miller, M.C.; Takahashi, S.; Seki, S.; Inose, S.; Yamamoto, T.

    1993-01-01

    A neutron coincidence counter has been designed for the measurement of plutonium powder contained in tall storage canisters. The counter was designed for installation in the Plutonium Fuel Production Facility fabrication plant. Each canister contains from one to five cans of PuO 2 . The neutron counter measures the spontaneous-fission rate from the plutonium and, when this is combined with the plutonium isotopic ratios, the plutonium mass is determined. The system can accommodate plutonium loadings up to 12 kg, with 10 kg being a typical loading. Software has been developed to permit the continuous operation of the system in an unattended mode. Authentication techniques have been developed for the system. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

  14. Application of plutonium inventory measurement system (PIMS) and temporary canister verification system (TCVS) at RRP

    International Nuclear Information System (INIS)

    Noguchi, Yoshihiko; Nakamura, Hironobu; Adachi, Hideto; Iwamoto, Tomonori

    2004-01-01

    In U-Pu co-denitration area at Rokkasho Reprocessing Plant (RRP), Plutonium Inventory Measurement System (PIMS) and Temporary Canister Verification System (TCVS) are installed to provide efficient and effective safeguards. PIMS measures Pu quantity inside pipes and vessels installed in glove boxes by total neutron counting method. PIMS consists of total 142 neutron detector attached on the wall and top of glove boxes and neutron count rates of each detectors are related to each other to calculate Pu quantity of each process areas. In this moment, inactive calibration using Cf-source was completed. On the other hand, TCVS measures Pu quantity of canisters inside temporary storage by coincidence counting method and it will be installed before the active test. These systems have monitoring function as additional measures. This paper describes specification, performance and measurement principles of PIMS and TCVS. (author)

  15. Economic, Energetic, and Environmental Performance of a Solar Powered Organic Rankine Cycle with Electric Energy Storage in Different Commercial Buildings

    Directory of Open Access Journals (Sweden)

    Emily Spayde

    2018-01-01

    Full Text Available This paper presents an analysis to determine the economic, energetic, and environmental benefits that could be obtained from the implementation of a combined solar-power organic Rankine cycle (ORC with electric energy storage (EES to supply electricity to several commercial buildings including a large office, a small office, and a full service restaurant. The operational strategy for the ORC-EES system consists in the ORC charging the EES when the irradiation level is sufficient to generate power, and the EES providing electricity to the building when there is not irradiation (i.e., during night time. Electricity is purchased from the utility grid unless it is provided by the EES. The potential of the proposed system to reduce primary energy consumption (PEC, carbon dioxide emission (CDE, and cost was evaluated. Furthermore, the available capital cost for a variable payback period for the ORC-EES system was determined for each of the evaluated buildings. The effect of the number of solar collectors on the performance of the ORC-EES is also studied. Results indicate that the proposed ORC-EES system is able to satisfy 11%, 13%, and 18% of the electrical demand for the large office, the small office and the restaurant, respectively.

  16. Further assessment studies of the Advanced Cold Process Canister

    International Nuclear Information System (INIS)

    Henshaw, J.; Hoch, A.; Sharland, S.M.

    1990-08-01

    A preliminary assessment of the performance of the Advanced Cold Process Canister (ACPC) was carried out recently by Marsh. The aim of the study presented in this report is to re-examine the validity of some of the assumptions made, and re-evaluate the canister performance as appropriate. Two areas were highlighted in the preliminary study as requiring more detailed quantitative evaluation. 1) Assessment of the risk of internal stress-corrosion cracking induced by irradiation of moist air inside the canister if, under fault conditions, significant water was carried into the canister before sealing. 2) Evaluation of the corrosion behaviour subsequent to first breach of outer container. (author)

  17. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2005-06-01

    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 the 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 100m/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. (orig.)

  18. 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.

  19. Storage of nuclear waste in long boreholes

    International Nuclear Information System (INIS)

    Sandstedt, H.; Wichmann, C.; Pusch, R.; Boergesson, L.; Loennerberg, B.

    1991-08-01

    This report constitutes a feasibility study for the storage of high level radioactive waste in long TBM drilled tunnels. The report will form the basis for a comparison with other concepts in future analysis of the isolation performance in a typical Swedish rock structure. The suggested repository concept consists of three parallel, 4.5 km long, horizontal tunnels at a depth of 500 m constructed using TBM technology. The tunnel diameter will be about 2.4 m for deployment of canisters with a diameter of 1.6 m. The space between the canisters and rock will be totally sealed off by bentonite. The study comprises the design of canisters, canister handling and deposition, near field design, near field sealing and behaviour, and technical design of the repository. The report also includes a tentative time schedule and cost estimate, incorporating the construction phase and deployment of canisters. (au)

  20. Key technical issues relating to safety of spent fuel dry storage in vaults: CASCAD system

    Energy Technology Data Exchange (ETDEWEB)

    Berge, F [Societe Generale pour les Techniques Nouvelles (SGN), 78 - Saint-Quentin-en-Yvelines (France)

    1994-12-31

    The operating CASCAD Facility at the Cadarashe site (FR) was commissioned in May 1990. Fuel is received in tight canisters which are transferred to storage pits in the vault and scheduled to be stored for up to 50 years. Canistering operations are performed in a cell of the reactor building.The paper describes the main functions of the facility as: cask receipt and shipping; fuel unloading; fuel conditioning; canisters emplacements in storage location; fuel storage; fuel retrieving and shipping at the end of the storage period; operation system and operation organization. Safety characteristics of the facility discussed are: fuel decay heat removal; subcriticality control and radiological protection. The fuel decay heat removal has two main purposes: (1) maintaining rod cladding temperature below a set limit in order to keep the fuel in its as received condition; (2) maintaining structures and equipment performing a safety function below the design temperature. The features of the sub-criticality control in the storage vault are such that sub-criticality in normal and accidental conditions is provided by the arrangement of pits in the vault. Radiological protection is based on limiting collective and individual annual dose equivalent to ALARA levels ensuring that they remain in any case below the set limits. Radiological protection system described consists in: confinement of radioactive materials for protection against its dissemination; radiation shielding for protection against irradiation. It is pointed out that all technical solutions presented are based on or adapted from proven technologies used in operating facilities in France or in other countries. The solution not only benefits from the experience of SGN in the design, construction and start-up of facilities for fuel or high level waste handling and storage, but also from the experience of the CEA and COGEMA groups in operating such facilities. 2 figs., 1 ref.

  1. Optimal design and operation of a thermal storage system for a chilled water plant serving pharmaceutical buildings

    Energy Technology Data Exchange (ETDEWEB)

    Henze, Gregor P. [University of Nebraska, Architectural Engineering, Omaha, NE 68182 (United States); Biffar, Bernd; Kohn, Dietmar [Boehringer Ingelheim Pharma GmbH and Co. KG, Biberach D-88400 (Germany); Becker, Martin P. [University of Applied Sciences Biberach, Architectural Engineering, Biberach D-88400 (Germany)

    2008-07-01

    A group of buildings in the pharmaceutical industry located in Southern Germany is experiencing a trend of growing cooling loads to be met by the chilled water plant composed of 10 chillers of greatly varying cost effectiveness. With a capacity shortfall inevitable, the question arises whether to install an additional chiller or improve the utilization of the existing chillers, in particular those with low operating costs per unit cooling, through the addition of a chilled water thermal energy storage (TES) system. To provide decision support in this matter, an optimization environment was developed and validated that adopts mixed integer programming as the approach to optimizing the chiller dispatch for any load condition, while an overarching dynamic programming based optimization approach optimizes the charge/discharge strategy of the TES system. In this fashion, the chilled water plant optimization is decoupled but embedded in the TES control optimization. The approach was selected to allow for arbitrary constraints and optimization horizons, while ensuring a global optimum to the problem. Optimization scenarios have been defined that include current load conditions as well cooling loads that are elevated by 25% from current conditions in order to reflect the expected growth in cooling demand in the near future; both scenarios analyzed the impact of storage capacity by investigating several TES tank capacities. The annual optimization runs revealed that - based on the elevated cooling load scenario - the smallest TES system pays back the incremental investment necessary for the TES system in about three years; based on today's cooling loads the static payback is approximately six years. As the efficiency and cost of operating the existing chillers vary over a wide range, the TES system allows for a reduction in operating costs for the chilled water plant by avoiding the operation of inefficient chillers (such as the single-stage absorption type) and

  2. Pitting corrosion on a copper canister

    International Nuclear Information System (INIS)

    Hermansson, H.P.; Beverskog, B.

    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

  3. Multi-Canister overpack pressure testing

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    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

  4. Building a columnar database on shared main memory-based storage

    OpenAIRE

    Tinnefeld, Christian

    2014-01-01

    In the field of disk-based parallel database management systems exists a great variety of solutions based on a shared-storage or a shared-nothing architecture. In contrast, main memory-based parallel database management systems are dominated solely by the shared-nothing approach as it preserves the in-memory performance advantage by processing data locally on each server. We argue that this unilateral development is going to cease due to the combination of the following three trends: a) Nowad...

  5. Annex D 200 Area Interim Storage Area Final Safety Analysis Report Volume 5 (FSAR) (Section 1 and 2)

    International Nuclear Information System (INIS)

    CARRELL, R.D.

    2003-01-01

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped offsite to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility (FFTF) Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the FFTF SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF)TRIGA--One Rad-Vault container stores two DOT-6M 3 containers and six NRF TRIGA casks. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-1 cask with an inner commercial light water reactor (LWR) canister, are used for storing commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

  6. Guidelines for developing efficient thermal conduction and storage models within building energy simulations

    International Nuclear Information System (INIS)

    Hillary, Jason; Walsh, Ed; Shah, Amip; Zhou, Rongliang; Walsh, Pat

    2017-01-01

    Improving building energy efficiency is of paramount importance due to the large proportion of energy consumed by thermal operations. Consequently, simulating a building's environment has gained popularity for assessing thermal comfort and design. The extended timeframes and large physical scales involved necessitate compact modelling approaches. The accuracy of such simulations is of chief concern, yet there is little guidance offered on achieving accurate solutions whilst mitigating prohibitive computational costs. Therefore, the present study addresses this deficit by providing clear guidance on discretisation levels required for achieving accurate but computationally inexpensive models. This is achieved by comparing numerical models of varying discretisation levels to benchmark analytical solutions with prediction accuracy assessed and reported in terms of governing dimensionless parameters, Biot and Fourier numbers, to ensure generality of findings. Furthermore, spatial and temporal discretisation errors are separated and assessed independently. Contour plots are presented to intuitively determine the optimal discretisation levels and time-steps required to achieve accurate thermal response predictions. Simulations derived from these contour plots were tested against various building conditions with excellent agreement observed throughout. Additionally, various scenarios are highlighted where the classical single lumped capacitance model can be applied for Biot numbers much greater than 0.1 without reducing accuracy. - Highlights: • Addressing the problems of inadequate discretisation within building energy models. • Accuracy of numerical models assessed against analytical solutions. • Fourier and Biot numbers used to provide generality of results for any material. • Contour plots offer intuitive way to interpret results for manual discretisation. • Results show proposed technique promising for automation of discretisation process.

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

    International Nuclear Information System (INIS)

    Makenas, B.J.

    1996-01-01

    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

  8. Risks in the transport and storage of liquefied natural gas. Sub-project 5-2: Investigation into building damage

    Science.gov (United States)

    Gouwens, C.; Dragosavic, M.

    The large reserves and increasing use of natural gas as a source of energy have resulted in its storage and transport becoming an urgent problem. Since a liquid of the same mass occupies only a fraction of the volume of a gas, it is economical to store natural gas as a liquid. Liquefied natural gas is stored in insulated tanks and also carried by ship at a temperature of -160 C to 170 C. If a serious accident allows the LNG to escape, a gas cloud forms. The results of a possible explosion from such a gas cloud are studied. The development of a leak, escape and evaporation, size and propagation of the gas cloud, the explosive pressures to be expected and the results on the environment are investigated. Damage to buildings is examined making use of the preliminary conclusions of the other sub-projects and especially the explosive pressures.

  9. Test manufacture of the canister insert 135

    International Nuclear Information System (INIS)

    Raiko, H.

    2005-10-01

    This report describes the insert-manufacturing test of a disposal canister for spent nuclear fuel that was made by Metso Foundries Jyvaeskylae Oy, in June 2004 on contract for Posiva Oy. The test manufacture was a part of the co-operation development programme of encapsulation technology between SKB AB and Posiva Oy. Insert casting was specified according to the current manufacturing specifications of SKB. The canister insert was of BWR-type with integral bottom. This was the second trial manufacture of this type of insert in Finland and, in total, the third test manufacture of insert by Metso Foundries Jyvaeskylae Oy. The result fulfilled all the requirements but the material mechanical properties of the cast material. The measured ultimate strength and elongation at rupture were lower than specified in the upper part of the cast. The reason for this was revealed in the metallurgical investigation of the cast material. The cast contained slag (dross). Avoiding the dross formation will be the most demanding challenge of the forthcoming development of the cast procedure. (orig.)

  10. Design analysis report for the canister

    International Nuclear Information System (INIS)

    Raiko, Heikki; Sandstroem, Rolf; Ryden, Haakan; Johansson, Magnus

    2010-04-01

    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

  11. Test manufacture of a canister insert

    International Nuclear Information System (INIS)

    Raiko, H.

    2004-11-01

    This report describes the insert-manufacturing test of a disposal canister for spent nuclear fuel that was made by Metso Paper Oy, Jyvaeskylae Foundry, in 2003 on contract for Posiva Oy. The test manufacture was a part of the co-operation development programme of encapsulation technology between SKB AB and Posiva Oy. Insert casting was specified according to the current manufacturing specifications of SKB. The canister insert was of BWR-type with integral bottom. This was the first trial manufacture of this type of insert in Finland and, in total, the second test manufacture of insert by Metso Paper. The result fulfilled all the requirements but the material mechanical properties and metallurgical structure of the cast material. The measured tensile strength, ultimate strength and elongation at rupture were lower than specified. The reason for this was revealed in the metallurgical investigation of the cast material. The nodulizing of the graphite was not occurred during the casting process according to the requirements. (orig.)

  12. Solid state phase change materials for thermal energy storage in passive solar heated buildings

    Science.gov (United States)

    Benson, D. K.; Christensen, C.

    1983-11-01

    A set of solid state phase change materials was evaluated for possible use in passive solar thermal energy storage systems. The most promising materials are organic solid solutions of pentaerythritol, pentaglycerine and neopentyl glycol. Solid solution mixtures of these compounds can be tailored so that they exhibit solid-to-solid phase transformations at any desired temperature within the range from less than 25 deg to 188 deg. Thermophysical properties such as thermal conductivity, density and volumetric expansion were measured. Computer simulations were used to predict the performance of various Trombe wall designs incorporating solid state phase change materials. Optimum performance was found to be sensitive to the choice of phase change temperatures and to the thermal conductivity of the phase change material. A molecular mechanism of the solid state phase transition is proposed and supported by infrared spectroscopic evidence.

  13. Construction of mixed waste storage RCRA facilities, Buildings 7668 and 7669: Environmental assessment

    International Nuclear Information System (INIS)

    1994-04-01

    The Department of Energy has prepared an environmental assessment, DOE/EA-0820, to assess the potential environmental impacts of constructing and operating two mixed waste Resource Conservation and Recovery Act (RCRA) storage facilities. The new facilities would be located inside and immediately west of the security-fenced area of the Oak Ridge National Laboratory Hazardous Waste Management Area in Melton Valley, Tennessee. Based on the analyses in the environmental assessment, the Department has determined that the proposed action does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act of 1969. Therefore, the preparation of an environmental impact statement is not required, and the Department is issuing this finding of no significant impact

  14. Retrospective dosimetry: dose evaluation using unheated and heated quartz from a radioactive waste storage building

    International Nuclear Information System (INIS)

    Jain, M.; Boetter-Jensen, L.; Murray, A.S.; Jungner, H.

    2002-01-01

    In the assessment of dose received from a nuclear accident, considerable attention has been paid to retrospective dosimetry using heated materials such as household ceramics and bricks. However, unheated materials such as mortar and concrete are more commonly found in industrial sites and particularly in nuclear installations. These materials contain natural dosemeters such as quartz, which usually is less sensitive than its heated counterpart. The potential of quartz extracted from mortar is a wall of a low-level radioactive-waste storage facility containing distributed sources of 60 Co and 13C s has been investigated. Dose-depth profiles based on small aliquots and single grains from the quartz extracted from the mortar samples are reported here. These are compared with results from heated quartz and polymineral fine grains extracted from an adjacent brick, and the integrated dose recorded by environmental TLDs. (author)

  15. Retrospective dosimetry: Dose evaluation using unheated and heated quartz from a radioactive waste storage building

    DEFF Research Database (Denmark)

    Jain, M.; Bøtter-Jensen, L.; Murray, A.S.

    2002-01-01

    In the assessment of dose received from a nuclear accident, considerable attention has been paid to retrospective dosimetry using heated materials such as household ceramics and bricks. However, unheated materials such as mortar and concrete are more commonly found in industrial sites......-137 has been investigated. Dose-depth profiles based on small aliquots and single grains from the quartz extracted from the mortar samples are reported here. These are compared with results from heated quartz and polymineral fine grains extracted from an adjacent brick, and the integrated dose...... and particularly in nuclear installations. These materials contain natural dosemeters Such as quartz. which usually is less sensitive than its heated counterpart. The potential of quartz extracted from mortar in a wall of a low-level radioactive-waste storage facility containing distributed sources of Co-60 and Cs...

  16. Current status of and problems in ice heat storage systems contributing to improving load rate. Air conditioning system utilizing ice heat storage and building frame storage (Takenaka Corporation); Fukaritsu kaizen ni kokensuru kori chikunetsu system no genjo to kadai. Kori chikunetsu to kutai chikunetsu wo riyoshita kucho system

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, Y.; Yoshitake, Y. [Takenaka Corp., Osaka (Japan)

    1998-02-01

    Development was made on a new air conditioning system, `building frame heat storage air conditioning system`, which combines an ice heat storage system with a building frame heat storage. With the building frame heat storage system, a damper is installed on an indoor device to blow cold air from the air conditioner onto slabs on the upper floor during nighttime power generating period. Heat is stored in beams, pillars and walls, and the shell absorbs and dissipates heat during daytime. Since general office buildings consume primary energy at 22.8% for heat source and 26.9% for transportation, which is greater, a natural coolant circulation type air conditioning system was developed as an air conditioning system for the secondary side. This made the building frame heat regeneration possible for the first time. With regard to heat storage quantity and heat dissipation quantity, the quantity of heat which can be stored during nighttime (10 hours) and dissipated during air conditioning using period in daytime (10 hours) is 80% of the stored heat quantity. This system was installed in a building in Kobe City. As a result of the measurement, it was found that indoor heat load reduction rate as a result of using the building frame heat storage was 24% or more in summer and 80% or more in winter. 7 figs., 2 tabs.

  17. Interim transfer canister for consolidating nuclear fuel rods

    International Nuclear Information System (INIS)

    Formanek, F.J.

    1987-01-01

    This patent describes a canister for receiving and consolidating a group of uniformly spaced apart nuclear fuel rods, comprising: a rectangular, vertically oriented straight back panel; a pair of oppositely disposed side panels connected perpendicularly to the back panel, having a vertical straight upper portion and an inwardly tapered lower portion; a front panel opposite the back panel and connected to the side panels, having a straight vertical upper portion and inwardly tapered lower portion; whereby the back, side and front panels define a rectangular upper opening at the upper end of the canister and a generally rectangular lower opening at the other end, the lower opening having a cross-sectional area less than one-half that of the upper opening; parallel plate members spanning the canister from the front panel to the back panel, each plate spaced from the other the same uniform distance, the plates extending downwardly into the tapered portion of the canister while remaining spaced above the tapered sidewalls; first base means at the lower end of the canister, removably mounted and having an oblique orientation generally downward from the front panel to the back panel, for guiding the fuel rods to be inserted preferentially toward the lower portion of the back panel; and second base means removably mounted within the canister below first base means and oriented transversely to the longitudinal extent of the canister, for supporting the fuel rods when the first base means is removed from the canister

  18. Design basis for the copper/steel canister

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    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

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

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

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

  20. User-Preference-Driven Model Predictive Control of Residential Building Loads and Battery Storage for Demand Response: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Xin [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Baker, Kyri A. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Christensen, Dane T. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Isley, Steven C. [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-08-21

    This paper presents a user-preference-driven home energy management system (HEMS) for demand response (DR) with residential building loads and battery storage. The HEMS is based on a multi-objective model predictive control algorithm, where the objectives include energy cost, thermal comfort, and carbon emission. A multi-criterion decision making method originating from social science is used to quickly determine user preferences based on a brief survey and derive the weights of different objectives used in the optimization process. Besides the residential appliances used in the traditional DR programs, a home battery system is integrated into the HEMS to improve the flexibility and reliability of the DR resources. Simulation studies have been performed on field data from a residential building stock data set. Appliance models and usage patterns were learned from the data to predict the DR resource availability. Results indicate the HEMS was able to provide a significant amount of load reduction with less than 20% prediction error in both heating and cooling cases.

  1. Electrochemical corrosion protection of storage water heaters in the building services; Elektrochemischer Korrosionsschutz von Speicher-Wassererwaermern in der Gebaeudetechnik

    Energy Technology Data Exchange (ETDEWEB)

    Bytyn, Wilfried [MAGONTEC GmbH, Bottrop (Germany)

    2012-06-15

    Storage water heaters currently experience a new consideration as a central thermal energy storage with an energy buffer characteristics. The contribution under consideration presents the principles and conditions of use for the cathodic corrosion protection of storage water heaters.

  2. FPSO - floating production storage and off loading -conversion or new build; FPSO - conversao ou construcao nova

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Fernando Sampaio; Mitidieri, Jorge Luiz

    2004-07-01

    The FPSO technology over the past years has rapidly evolved from tanker based units with classical mooring system to purpose built vessels equipped with dynamic positioning assistance, sophisticated permanent turrets or disconnectable mooring systems. The growth in the number of FPSO's in the 1980's was largely based converted tankers but new-builds now represent 40% of the FPSO fleet. This paper intends to discuss the main technical and economical aspects to be considered when selecting the project for a FPSO construction, including, among others, general and structural conditions for the takers to be converted, safety considerations, systems to be installed and general analysis of engineering design. The study compares construction schedule and the time schedule influence on project conclusion. The evaluation of the designed life time of each vessel an its implications on the project cost and the importance on the decision is also evaluated. (author)

  3. FPSO - floating production storage and off loading -conversion or new build; FPSO - conversao ou construcao nova

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Fernando Sampaio; Mitidieri, Jorge Luiz

    2004-07-01

    The FPSO technology over the past years has rapidly evolved from tanker based units with classical mooring system to purpose built vessels equipped with dynamic positioning assistance, sophisticated permanent turrets or disconnectable mooring systems. The growth in the number of FPSO's in the 1980's was largely based converted tankers but new-builds now represent 40% of the FPSO fleet. This paper intends to discuss the main technical and economical aspects to be considered when selecting the project for a FPSO construction, including, among others, general and structural conditions for the takers to be converted, safety considerations, systems to be installed and general analysis of engineering design. The study compares construction schedule and the time schedule influence on project conclusion. The evaluation of the designed life time of each vessel an its implications on the project cost and the importance on the decision is also evaluated. (author)

  4. Characterization of Emissions from Building Products: Selection Criteria, Storage and Handling of Indoor Materials for Experiments

    DEFF Research Database (Denmark)

    Kjær, Ulla Dorte; Nielsen, Peter A.; Marcussen, Lis

    1996-01-01

    In research projects it is usually impossible to study more than a very limited aount of different products. Care and consideration must therefore be taken in the selection. The following criteria were used to select five products for chemical and sensory emission testing: The products should...... be indoor climate relevant, meaning that they should be in direct contact with indoor air when used in buildings, and their emission should contribute significantly to the indoor air concentrations for instance due to a large surface area. They should be representative in terms of being widely used, now...... that the emission mechanisms do not change as a consequence of physical processes or chemical reactions. The five products selected were PVC flooring, waterborne varnish on beechwood parquet, carpet, sealant, and waterborne paint on gypsum board. The products were brought to the laboratory immediately after...

  5. Considerations for Disposition of Dry Cask Storage System Materials at End of Storage System Life

    International Nuclear Information System (INIS)

    Howard, Rob; Van den Akker, Bret

    2014-01-01

    Dry cask storage systems are deployed at nuclear power plants for used nuclear fuel (UNF) storage when spent fuel pools reach their storage capacity and/or the plants are decommissioned. An important waste and materials disposition consideration arising from the increasing use of these systems is the management of the dry cask storage systems' materials after the UNF proceeds to disposition. Thermal analyses of repository design concepts currently under consideration internationally indicate that waste package sizes for the geologic media under consideration may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded into the dry storage canisters currently in use. In the United States, there are already over 1650 of these dry storage canisters deployed and approximately 200 canisters per year are being loaded at the current fleet of commercial nuclear power plants. There is about 10 cubic meters of material from each dry storage canister system that will need to be dispositioned. The concrete horizontal storage modules or vertical storage overpacks will need to be reused, re-purposed, recycled, or disposed of in some manner. The empty metal storage canister/cask would also have to be cleaned, and decontaminated for possible reuse or recycling or disposed of, likely as low-level radioactive waste. These material disposition options can have impacts of the overall used fuel management system costs. This paper will identify and explore some of the technical and interface considerations associated with managing the dry cask storage system materials. (authors)

  6. Dry-type radioactive material storage facility

    International Nuclear Information System (INIS)

    Yamanaka, Yasuharu; Matsuda, Masami; Kanai, Hidetoshi; Ganda, Takao.

    1996-01-01

    A plurality of container tubes containing a plurality of canisters therein are disposed in a canister storage chamber. High level radioactive materials are filled in the canisters in the form of glass solidification materials. The canister storage chamber is divided into two cooling channels by a horizontal partition wall. Each of the container tubes is suspended from a ceiling slab and pass through the horizontal partition wall. Namely, each of the container tubes vertically traverses the cooling channel formed between the ceiling slab and the partition wall and extends to the cooling channel formed between the partition wall and a floor slab. Cooling gases heated in the cooling channel below the partition wall are suppressed from rising to the cooling channel above the partition wall. Therefore, the container tubes are efficiently cooled even in a cooling channel above the partition wall to unify temperature distribution in the axial direction of the container tubes. (I.N.)

  7. Dry storage of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Tolmie, R.D.

    1983-01-01

    In transferring radioactive material between the preparation and clean chambers of a dry storage complex, irradiated nuclear fuel is posted from the preparation chamber to a sealable canister supported in a closable bucket in the clean chamber, or a contaminated sealed canister is posted from a closed bucket in the clean chamber into the preparation chamber by using a facility comprising two coaxial tubes constituting a closable orifice between the two chambers, the tubes providing sealing means for the bucket, and masking means for the bucket and canister closures together with means for withdrawing the closures into the preparation chamber. (author)

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

    International Nuclear Information System (INIS)

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

    1994-04-01

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

  9. Selection and evaluation of inner material candidates for Spanish high level radioactive waste canisters

    International Nuclear Information System (INIS)

    Puig, Francesc; Dies, Javier; Sevilla, Manuel; Pablo, Joan de; Pueyo, Juan Jose; Miralles, Lourdes; Martinez-Esparza, Aurora

    2007-01-01

    This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose. (authors)

  10. Carbon Nanotube Thin Films for Active Noise Cancellation, Solar Energy Harvesting, and Energy Storage in Building Windows

    Science.gov (United States)

    Hu, Shan

    This research explores the application of carbon nanotube (CNT) films for active noise cancellation, solar energy harvesting and energy storage in building windows. The CNT-based components developed herein can be integrated into a solar-powered active noise control system for a building window. First, the use of a transparent acoustic transducer as both an invisible speaker for auxiliary audio playback and for active noise cancellation is accomplished in this work. Several challenges related to active noise cancellation in the window are addressed. These include secondary path estimation and directional cancellation of noise so as to preserve auxiliary audio and internal sounds while preventing transmission of external noise into the building. Solar energy can be harvested at a low rate of power over long durations while acoustic sound cancellation requires short durations of high power. A supercapacitor based energy storage system is therefore considered for the window. Using CNTs as electrode materials, two generations of flexible, thin, and fully solid-state supercapacitors are developed that can be integrated into the window frame. Both generations consist of carbon nanotube films coated on supporting substrates as electrodes and a solid-state polymer gel layer for the electrolyte. The first generation is a single-cell parallel-plate supercapacitor with a working voltage of 3 Volts. Its energy density is competitive with commercially available supercapacitors (which use liquid electrolyte). For many applications that will require higher working voltage, the second-generation multi-cell supercapacitor is developed. A six-cell device with a working voltage as high as 12 Volts is demonstrated here. Unlike the first generation's 3D structure, the second generation has a novel planar (2D) architecture, which makes it easy to integrate multiple cells into a thin and flexible supercapacitor. The multi-cell planar supercapacitor has energy density exceeding that of

  11. Influence of internal thermal mass on the indoor thermal dynamics and integration of phase change materials in furniture for building energy storage

    DEFF Research Database (Denmark)

    Johra, Hicham; Heiselberg, Per Kvols

    2017-01-01

    The increasing share of intermittent renewable energy on the grid encourages researchers to develop demand-side management strategies. Passive heat storage in the indoor space is a promising solution to improve the building energy flexibility. It relies on an accurate control of the transient...... building temperature. However, many of the current numerical models for building energy systems assume empty rooms and do not account entirely for the internal thermal inertia of objects like furniture. This review article points out that such assumption is not valid for dynamic calculations...

  12. Multi-Canister overpack sealing configuration

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The Spent Nuclear Fuel (SNF) position regarding the Multi-Canister Overpack (MCO) sealing configuration is to initially rely on an American Society of Mechanical Engineers (ASME) Section III Subsection NB code compliant mechanical closure/sealing system to quickly and safely establish and maintain full confinement of radioactive materials prior to and during MCO fuel drying activities. Previous studies have shown the mechanical seal to be the preferred closure method, based on dose, cost, and schedule considerations. The cost and schedule impacts of redesigning the mechanical closure to a welded shield plug do not support changing the closure system. The SNF Project has determined that the combined mechanical/welded closure system meets or exceeds the regulatory requirements to provide redundant seals while accommodating key safety and schedule limitations that are unique to K Basins fuel removal effort

  13. Multi-Canister overpack internal HEPA filters

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    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

  14. Thermal performance of a concrete cask: Methodology to model helium leakage from the steel canister

    International Nuclear Information System (INIS)

    Penalva, J.; Feria, F.; Herranz, L.E.

    2017-01-01

    Highlights: • A thermal analysis of the canister during a loss of leaktightness has been performed. • Methodologies that predict fuel temperatures and heat up rates have been developed. • Casks with heat loads below 20 kW would never exceed the thermal threshold. - Abstract: Concrete cask storage systems used in dry storage allocate spent fuel within containers that are usually filled with helium at a certain pressure. Potential leaks from the container would result in a cooling degradation of fuel that might jeopardize fuel integrity if temperature exceeded a threshold value. According to ISG-11, temperatures below 673 K ensure fuel integrity preservation. Therefore, the container thermal response to a loss of leaktightness is of utmost importance in terms of safety. In this work, a thermo-fluid dynamic analysis of the canister during a loss of leaktightness has been performed. To do so, steady-state and transient Computational Fluid Dynamics (CFD) simulations have been carried out. Likewise, it has been developed two methodologies capable of estimating peak fuel temperatures and heat up rates resulting from a postulated depressurization in a dry storage cask. One methodology is based on control theory and transfers functions, and the other methodology is based on a linear relationship between the inner pressure and the maximum temperature. Both methodologies have been verified through comparisons with CFD calculations. The period of time to achieve the temperature threshold (673 K) is a function of pressure loss rate and decay heat of the fuel stored in the container; in case of a fuel canister with 30 kW the period of time to reach the thermal limit takes between half day (fast pressure loss) and one week (slow pressure loss). In case of a 15% reduction of the decay heat, the period of time to achieve the thermal limit increase up to a few weeks. The results highlight that casks with heat loads below 20 kW would never exceed the thermal threshold (673 K).

  15. 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

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

    International Nuclear Information System (INIS)

    Slate, S.C.; Pitman, S.G.; Nesbitt, J.F.; Partain, W.L.

    1982-08-01

    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

  17. Thermal energy storage with phase change materials (PCMs) for the improvement of the energy performance of buildings

    Science.gov (United States)

    Soares, Nelson

    The improvement of the energy efficiency of buildings during their operational phase is an active area of research. The markets are looking for new technologies, namely new thermal energy storage (TES) systems, which can be used to reduce buildings' dependency on fossil fuels, to make use of renewable energy sources and to contribute to match energy supply and demand efficiently. The main goals of this thesis are: (i) to evaluate the heat transfer with solid-liquid phase-change through small TES units filled with phase-change materials (PCMs), providing experimental data to be used in the design of new TES systems for buildings and in the validation of numerical models, and (ii) to provide some guidelines for the incorporation of PCM-drywalls in buildings aiming to reduce the energy demand for heating and cooling by making use of the latent heat from the phase-change processes of PCMs. The first part of this thesis refers to the experimental study of the heat transfer through a vertical stack of metallic rectangular cavities filled with different PCMs (a microencapsulated and a free-form PCM). The research carried out aims: (i) to analyze the melting and solidification processes of the PCM within the enclosures, (ii) to evaluate the influence of the aspect ratio of the cavities on the heat transfer and (iii) to discuss which type of PCM is better for specific cases. As a result, a big amount of experimental data for benchmarking and validation of numerical models is made available to the scientific community. Moreover, the results allow discussing which arrangement of the TES unit is better for specific applications considering the thermal regulation effect during charging, the influence of subcooling during discharging, and the influence of natural convection during both processes. It is shown that the effect of natural convection in the free-form PCM must be considered in any simulation to better describe the charging process. During discharging, subcooling must

  18. 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

  19. MANAGING THE LOAD SCHEDULE OF THE ADMINISTRATIVE BUILDING TAKING INTO ACCOUNT EMERGING RISKS WHEN CONNECTING THE KINETIC ENERGY STORAGE TO THE POWER SUPPLY SYSTEM

    Directory of Open Access Journals (Sweden)

    S. Yu. Shevchenko

    2017-12-01

    Full Text Available Purpose. The purpose of the paper is to analyze load schedules of the administrative building and develop a structural scheme for connecting the kinetic energy storage in the power supply system of this building, which will allow using it as a consumer regulator, as well as a theoretical study of the risks that arise. Methodology. To conduct the research, the theory of designing internal electrical networks of buildings, the theory of plotting electric load graphs, methods of the theory of electromechanical systems and for analyzing the risk system, the T. Saati method of hierarchies were used. Results. The structure of kinetic energy storage (KES connection to the power supply system of the administrative building is developed and the structural diagram of the KES proposed for installation is given, the average daily winter and summer load schedules are presented, a set of groups and subgroups of risks and their influence on the work of the power supply system of the building are connected with the connection of the KES. Originality. For the first time, the application of the kinetic energy storage in the power supply system of the building with the analysis of emerging risks is considered, which makes it possible to improve the reliability of the developed system and the efficiency of load regulation. Practical value. The application of the proposed scheme will make it possible to use administrative buildings as load regulators of the external power supply system, and also effectively manage the load in the internal power supply system of the building.

  20. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2003-01-01

    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.)

  1. The Swedish Concept for Disposal of Spent Nuclear Fuel: Differences Between Vertical and Horizontal Waste Canister Emplacement

    International Nuclear Information System (INIS)

    Bennett, D.G.; Hicks, T.W.

    2005-10-01

    The Swedish Nuclear Power Inspectorate (SKI) is preparing for the review of licence applications related to the disposal of spent nuclear fuel. The Swedish Nuclear Fuel and Waste Management Company (SKB) refers to its proposals for the disposal of spent nuclear fuel as the KBS-3 concept. In the KBS-3 concept, SKB plans that, after 30 to 40 years of interim storage, spent fuel will be disposed of at a depth of about 500 m in crystalline bedrock, surrounded by a system of engineered barriers. The principle barrier to radionuclide release is a cylindrical copper canister. Within the copper canister, the spent fuel is supported by a cast iron insert. Outside the copper canister is a layer of bentonite clay, known as the buffer, which is designed to provide mechanical protection for the canisters and to limit the access of groundwater and corrosive substances to their surfaces. The bentonite buffer is also designed to sorb radionuclides released from the canisters, and to filter any colloids that may form within the waste. SKB is expected to base its forthcoming licence applications on a repository design in which the waste canisters are emplaced in vertical boreholes (KBS-3V). However, SKB has also indicated that it might be possible and, in some respects, beneficial to dispose of the waste canisters in horizontal tunnels (KBS-3H). There are many similarities between the KBS-3V and KBS-3H designs. There are, however, uncertainties associated with both of the designs and, when compared, both possess relative advantages and disadvantages. SKB has identified many of the key factors that will determine the evolution of a KBS-3H repository and has plans for research and development work in many of the areas where the differences between the KBS-3V and KBS-3H designs mean that they could be significant in terms of repository performance. With respect to the KBS-3H design, key technical issues are associated with: 1. The accuracy of deposition drift construction. 2. Water

  2. National waste terminal storage conceptual reference repository description

    International Nuclear Information System (INIS)

    Odgers, I.L.; Collings, J.L.

    1981-01-01

    The conceptual reference repository description (CRRD) discusses, from a conceptual engineering standpoint, the structures, systems, equipment, and operations necessary to: (1) receive unreprocessed spent fuel assemblies in standard casks; (2) unload these assemblies; (3) place them in canisters; (4) transport the canisters to underground storage locations in the salt dome; and (5) place the canisters in terminal storage. The CRRD also elaborates on the concepts for retrieval and recovery of the spent fuel after burial and describes the development of the shafts and the underground areas, as well as the supporting operational utility and administrative features of the repository

  3. A simplified model to study the location impact of latent thermal energy storage in building cooling heating and power system

    International Nuclear Information System (INIS)

    Zhang, Yin; Wang, Xin; Zhang, Yinping; Zhuo, Siwen

    2016-01-01

    Introducing the thermal energy storage (TES) equipment into the building cooling heating and power (BCHP) system proves to be an effective way to improve the part load performance of the whole system and save the primary energy consumption. The location of TES in BCHP has a great impact on the thermal performance of the whole system. In this paper, a simplified model of TES-BCHP system composed of a gas turbine, an absorption chiller/an absorption heat pump, and TES equipment with phase change materials (PCM) is presented. In order to minimize the primary energy consumption, the performances of BCHP systems with different PCM-TES locations (upstream and downstream) are analyzed and compared, for a typical hotel and an office building respectively. Moreover, the influence of the thermal performance of PCM-TES equipment on the energy saving effect of the whole system is investigated. The results confirm that PCM-TES can improve the energy efficiency and reduce the installed capacities of energy supply equipment, and that the optimal TES location in BCHP highly depends on the thermal performance of the TES equipment and the user load characteristics. It also indicates that: 1) the primary energy saving ratio of PCM-TES-BCHP increases with increasing NTU of TES; 2) for the studied cases, downstream TES location becomes more preferable when user loads fluctuate greatly; 3) only downstream TES can reduce the installed capacities of absorption chiller/absorption heat pump. This work can provide guidance for PCM-TES-BCHP system design. - Highlights: • A simplified model of the PCM-TES-BCHP system is established. • TES can increase energy efficiency and decrease installed capacity of equipment. • Primary energy saving ratio increases with increasing NTU of TES. • Downstream TES location is more preferable when user loads fluctuate greatly. • Optimal TES location depends on equipment performances and load characteristics.

  4. DOE requests waiver on double containment for HLW canisters

    International Nuclear Information System (INIS)

    Lobsenz, G.

    1994-01-01

    The Energy Department has asked the Nuclear Regulatory Commission to waive double containment requirements for vitrified high-level radioactive waste canisters, saying the additional protection is not necessary and too costly. NRC said it had received a petition from DOE contending that the vitrified waste canisters were durable enough without double containment to prevent any potential plutonium release during handling and shipping. DOE said testing had shown that the vitrified waste canisters were similar - even superior - in durability to spent reactor fuel shipments, which NRC specifically exempted from the double containment requirement

  5. 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

  6. Comprehensive work plan for Building 3001 storage canal at the Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1997-01-01

    This Comprehensive Work Plan describes the method of accomplishment to replace the shielding protection of the water in the canal with a controlled low strength material (CLSM) 4. The canal was used during the operation of the Oak Ridge Graphite Reactor in the 1940s and 1950s to transport spent fuel slugs and irradiated test materials from the reactor, under water to the hot cell in Building 3019 for further processing, packaging, and handling. After the reactor was shut down, the canal was used until 1990 to store some irradiated materials until they could be transferred to a Solid Waste Storage Area. This task has the following objectives and components: (1) minimize potential future risk to human health and the environment; (2) reduce surveillance and maintenance cost of the canal; (3) perform site preparation activities; (4) replace the water in the canal with a solid CLSM; (5) pump the water to the Process Waste Treatment System (PWTS) for further processing at the same rate that the CLSM is pumped under the water; (6) remove the water using a process that will protect the workers and the public in the visitors area from contamination while the CLSM is being pumped underneath the water; (7) painting a protective coating material over the CLSM after the CLSM has cured

  7. Comments on 'SKB RD and D-Programme 98'. Focused on canister integrity and corrosion

    International Nuclear Information System (INIS)

    Bowyer, W.H.; Hermansson, H.P.

    1999-04-01

    According to the Act on Nuclear Activities the nuclear utilities are requested to submit a comprehensive programme for research and development every third year, aiming at the safe storage of radioactive waste produced by the nuclear power plants. The latest was published by SKB in September 1998 and is called the 'RDandD Programme 98'. The work presented in the present report was commissioned by SKI and is a result of reading the 'RDandD Programme 98' and related reports with focus on canister production, integrity and corrosion. We find that those parts of the programme often are difficult to follow owing to the lack of detail in the Programme and in one of the supporting reports. In our opinion this will make the work difficult to monitor by SKI and SKB. We also feel that the interpretation of information already available often is overoptimistic. As a consequence the difficulties ahead are understated and the programme is allowed to converge too quickly. We agree that the materials choices for both the inner and outer canisters are appropriate providing they both can be produced commercially and in a satisfactory metallurgical condition, that they can be quality assured and that no further unforeseen difficulties arise. We also agree that alternative technologies merit consideration for production of the outer canister and that alternative joining processes should be studied. We are actually concerned that greater prominence is not given to the alternatives in the programme. 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 time-scale 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

  8. Zero-Headspace Coal-Core Gas Desorption Canister, Revised Desorption Data Analysis Spreadsheets and a Dry Canister Heating System

    Science.gov (United States)

    Barker, Charles E.; Dallegge, Todd A.

    2005-01-01

    Coal desorption techniques typically use the U.S. Bureau of Mines (USBM) canister-desorption method as described by Diamond and Levine (1981), Close and Erwin (1989), Ryan and Dawson (1993), McLennan and others (1994), Mavor and Nelson (1997) and Diamond and Schatzel (1998). However, the coal desorption canister designs historically used with this method have an inherent flaw that allows a significant gas-filled headspace bubble to remain in the canister that later has to be compensated for by correcting the measured desorbed gas volume with a mathematical headspace volume correction (McLennan and others, 1994; Mavor and Nelson, 1997).

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

    International Nuclear Information System (INIS)

    Hoekmark, Harald; Faelth, Billy

    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

  10. Vacuum Drying Tests for Storage of Aluminum Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Chen, K.F.; Large, W.S.; Sindelar, R.L.

    1998-05-01

    A total inventory of up to approximately 32,000 aluminum-based spent nuclear fuel (Al SNF) assemblies are expected to be shipped to Savannah River Site (SRS) from domestic and foreign research reactors over the next several decades. Treatment technologies are being developed as alternatives to processing for the ultimate disposition of Al SNF in the geologic repository. One technology, called Direct/Co-disposal of Al SNF, would place the SNF into a canister ready for disposal in a waste package, with or without canisters containing high-level radioactive waste glass logs, in the repository. The Al SNF would be transferred from wet storage and would need to be dried in the Al SNF canister. The moisture content inside the Al SNF canister is limited to avoid excessive Al SNF corrosion and hydrogen buildup during interim storage before disposal. A vacuum drying process was proposed to dry the Al SNF in a canister. There are two major concerns for the vacuum drying process. One is water inside the canister could become frozen during the vacuum drying process and the other one is the detection of dryness inside the canister. To vacuum dry an irradiated fuel in a heavily shielded canister, it would be very difficult to open the lid to inspect the dryness during the vacuum drying operation. A vacuum drying test program using a mock SNF assembly was conducted to demonstrate feasibility of drying the Al SNF in a canister. These tests also served as a check-out of the drying apparatus for future tests in which irradiated fuel would be loaded into a canister under water followed by drying for storage

  11. MCC-15: waste/canister accident testing and analysis method

    International Nuclear Information System (INIS)

    Slate, S.C.; Pulsipher, B.A.; Scott, P.A.

    1985-02-01

    The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory (PNL) is developing standard tests to characterize the performance of nuclear waste forms under normal and accident conditions. As part of this effort, the MCC is developing MCC-15, Waste/Canister Accident Testing and Analysis. MCC-15 is used to test canisters containing simulated waste forms to provide data on the effects of accidental impacts on the waste form particle size and on canister integrity. The data is used to support the design of transportation and handling equipment and to demonstrate compliance with repository waste acceptance specifications. This paper reviews the requirements that led to the development of MCC-15, describes the test method itself, and presents some early results from tests on canisters representative of those proposed for the Defense Waste Processing Facility (DWPF). 13 references, 6 figures

  12. Multi-canister overpack operations and maintenance manual

    International Nuclear Information System (INIS)

    PIERCE, S.R.

    1999-01-01

    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

  13. Grain boundary corrosion of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes

    2006-01-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository

  14. Characterization of materials for waste-canister compatibility studies

    International Nuclear Information System (INIS)

    McCoy, H.E.; Mack, J.E.

    1981-10-01

    Sample materials of 7 waste forms and 15 potential canister materials were procured for compatibility tests. These materials were characterized before being placed in test, and the results are the main topic of this report. A test capsule was designed for the tests in which disks of a single waste form were contacted with duplicate samples of canister materials. The capsules are undergoing short-term tests at 800 0 C and long-term tests at 100 and 300 0 C

  15. Grain boundary corrosion of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes [Corrosion and Metals Research Inst. (KIMAB), Stockholm (Sweden)

    2006-01-15

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository.

  16. Extended used Fuel Storage: EPRI Perspective and Collaboration Initiatives

    International Nuclear Information System (INIS)

    Kessler, John; Waldrop, Keith

    2014-01-01

    This paper describes three main activities the Electric Power Research Institute (EPRI) is undertaking to establish the technical bases for extended (long-term) storage: the Extended Storage Collaboration Program (ESCP); inspection of stainless steel (SS) used fuel dry storage canisters currently in service; and a proposed data collection from a full-scale, bolted lid, metal cask containing high burnup (>45 GWd/MTU) used fuel (the 'Demo'). ESCP is a voluntary organization focused on information sharing and providing the opportunity for more formal collaboration. The SS canister inspection program involves visual examination, canister surface temperature measurements, and collection of contaminants accumulating on the canister surfaces during operation. The Demo program involves the use of a specially instrumented lid allowing for the introduction of thermocouples inside the loaded cask as was as providing the ability to collect cask cavity gas samples. (authors)

  17. QA/QC For Radon Concentration Measurement With Charcoal Canister

    International Nuclear Information System (INIS)

    Pantelic, G.; Zivanovic, M.; Rajacic, M.; Krneta Nikolic, J.; Todorovic, D.

    2015-01-01

    The primary concern of any measuring of radon or radon progeny must be the quality of the results. A good quality assurance program, when properly designed and diligently followed, ensures that laboratory staff will be able to produce the type and quality of measurement results which is needed and expected. 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. Upon closing the detectors, the measurement was carried out after achieving the equilibrium between radon and its daughters (at least 3 hours) using NaI or HPGe detector. Radon concentrations as well as measurement uncertainties were calculated according to US EPA protocol 520/5-87-005. Detectors used for the measurements were calibrated by 226Ra standard of known activity in the same geometry. Standard and background canisters are used for QA and QC, as well as for the calibration of the measurement equipment. Standard canister is a sealed canister with the same matrix and geometry as the canisters used for measurements, but with the known activity of radon. Background canister is a regular radon measurement canister, which has never been exposed. The detector background and detector efficiency are measured to ascertain whether they are within the warning and acceptance limits. (author).

  18. Drop tests of the Three Mile Island knockout canister

    International Nuclear Information System (INIS)

    Box, W.D.; Aaron, W.S.; Shappert, L.B.; Childress, P.C.; Quinn, G.J.; Smith, J.V.

    1986-09-01

    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

  19. Research on corrosion aspects of the advanced cold process canister

    International Nuclear Information System (INIS)

    Blackwood, D.J.; Hoch, A.R.; Naish, C.C.; Rance, A.

    1994-01-01

    The Advanced Cold Process Canister (ACPC) is a waste canister being developed jointly by SKB and TVO for the disposal of spent nuclear fuel. It comprises an outer copper canister, with a carbon steel canister inside. A concern regarding the use of the ACPC is that, in the unlikely event that the outer copper canister is penetrated, the anaerobic corrosion of the carbon steel container may result in the formation of hydrogen gas bubbles. These bubbles could disrupt the backfill, and thus increase water flow through the near field and the flux of radionuclides to the host geology. A number of factors that influence the rate at which hydrogen evolves as a result of the anaerobic corrosion of carbon steel in artificial granitic groundwaters have been investigated. A previously observed, time-dependent decline in the hydrogen evolution rate has been confirmed as being due to the production of magnetite film. Once the magnetite film is about 0.7-1.0 μm thick, the rate of hydrogen evolution reaches a steady state value. The pH and the ionic strength of the groundwater were both found to influence the long-term hydrogen evolution rate. The results of the experimental programme were used to update a model of the corrosion behaviour and hydrogen production from the Advanced Cold Process Canister. 36 figs, 5 tabs, 13 refs

  20. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

    1999-01-01

    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)

  1. SiteChar. Characterisation of European CO2 storage. Deliverable D8.2. Trust building and raising public awareness

    Energy Technology Data Exchange (ETDEWEB)

    Brunsting, S.; Pol, M.; Mastop, E.A. [ECN Policy Studies, Energy research Centre of the Netherlands ECN, Amsterdam (Netherlands); Kaiser, M.; Zimmer, R. [Unabhaengiges Institut fuer Umweltfragen UfU, Berlin (Germany); Shackley, S.; Mabon, L.; Howell, R. [Scottish Carbon Capture and Storage SCCS, Edinburg, Scotland (United Kingdom)

    2012-08-15

    At local level, public support has proven crucial to the implementation of CO2 capture and storage (CCS) demonstration projects. Whereas no method exists to guarantee public acceptability of any project, a constructive stakeholder and community engagement process does increase the likelihood thereof. This deliverable is a follow-up to deliverable D8.1 'Social site characterisation'. Social site characterisation can be used as an instrument to explore, plan and evaluate a process of active and constructive local stakeholder and citizen engagement in a prospective CCS project as a parallel activity to technical site characterisation. It serves as an analytical tool to describe the local social circumstances in the area and to design and evaluate stakeholder and community engagement efforts with the aims of building trust and raising public awareness. Using results from the social site characterisation of the area, the present deliverable focuses on the second purpose. It presents results from public engagement activities designed to raise public awareness and inform public opinion of a prospective CCS site in Poland (onshore) and the UK (offshore): focus conferences. Furthermore, by initiating an enhanced cooperation in planning of new storage sites between project developers, authorities and the local public, focus conferences aim to serve as a 'hinge' between social site characterisation as a research effort and application to real-life project settings. The focus conferences are part of a range of public engagement activities including the setup of public information websites on generic and site-specific CCS, information meetings. A second survey eventually shall evaluate the results of the public engagement activities. The aim of the focus conferences was to raise public awareness and assist public opinion forming processes of a prospective CCS site in Poland (onshore) and the UK (offshore). At the same time, it aimed to present and test a

  2. White Paper: Multi-purpose canister (MPC) for DOE-owned spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    Knecht, D.A.

    1994-04-01

    The paper examines the issue, What are the advantages, disadvantages, and other considerations for using the MPC concept as part of the strategy for interim storage and disposal of DOE-owned SNF? The paper is based in part on the results of an evaluation made for the DOE National Spent Fuel Program by the Waste Form Barrier/Canister Team, which is composed of knowledgeable DOE and DOE-contractor personnel. The paper reviews the MPC and DOE SNF status, provides criteria and other considerations applicable to the issue, and presents an evaluation, conclusions, and recommendations. The primary conclusion is that while most of DOE SNF is not currently sufficiently characterized to be sealed into an MPC, the advantages of standardized packages in handling, reduced radiation exposure, and improved human factors should be considered in DOE SNF program planning. While the design of MPCs for DOE SNF are likely premature at this time, the use of canisters should be considered which are consistent with interim storage options and the MPC design envelope

  3. Monitoring the Sanation Works in the Building Used as a Temporary Radioactive Material Storage at the Ruder Boskovic Institute in Zagreb

    International Nuclear Information System (INIS)

    Prlic, I.; Medakovic, S.; Novakovic, M.

    2016-01-01

    The sanation works regarding radioactive waste presently stored in the TSRM building at the Rudjer Boskovic Institute (RBI) formally ended in its first phase, by sealing the entrance into the TSRM entitled: old storage - area 1 (on the left) and new storage - area 2 (on the right) on the ground floor of the TSRM building at the RBI on 6th August 2015 After that day IMI performed the ambient dose equivalent rate (Hx(10)) measurements - radiological monitoring of the entire area around the TSRM RBI site (Figure 2) and additionally, of the area in front of the TSRM RBI building. Measurements were performed on an identical area as on 7th July 2015, when the initial - measurement 'zero' - of the H*(10) was performed. Measurements were done to establish the initial burden of the sanation site (Figure 2) and its environment from ionising radiation originating from radioactive sources stored within the TSRM RBI and disposed in its immediate environment. Since it was not stipulated whether environmental (indoor) dosimetry data are also necessary for the H*(10) inside the storage space (old and new storage) of the PSRM, only data for the exterior, i.e. environment of the PSRM RBI are represented here. Most of these data are also comprised in the Final Report of the Project Task IMI Class 07-75/15-0143-2/2; Reg.no.:100-08/15-1620. It should be noted that, on the day of the beginning of the renovation-sanation works inside the TSRM RBI, radioactive material was also found outside the TSRM building (this is discussed in the Final Report of the Project Task IMI Class 07-75/15-0143-2/2; Reg.no.:100-08/15-1620). (author).

  4. Corrosion studies on HGW-canister materials for marine disposal

    International Nuclear Information System (INIS)

    Taylor, K.J.; Bland, I.D.; Smith, S.; Marsh, G.P.

    1986-03-01

    Results are presented from theoretical and experimental work undertaken to investigate and assess the general corrosion behaviour of carbon steel canister/overpacks for heat generating nuclear waste under marine disposal conditions. The mean general corrosion rates of carbon steels, determined experimentally by polarisation resistance measurements on specimens in on-going immersion tests, are between 65-124 μm yr -1 at 90 0 C and 5-25 μm yr -1 at 25 0 C and are tending to increase with time. Anomalously high corrosion rates are being indicated by similar tests at 50 0 C. It is not clear what reliance should be placed on the polarisation resistance results, however, and therefore no conclusion will be drawn until the tests are dismantled and inspected in the 1985/86 programme. Tests with γ-radiation on forged carbon steel specimens immersed in deaerated seawater at 90 0 C show that this causes an acceleration of corrosion rate at the three dose rates down to at least 300 R h -1 . Deep ocean sediment from GME also accelerates the corrosion rate of carbon steel in deaerated seawater both with and without γ-radiation. The effect diminishes with continued exposure and is thought to be due to the presence of either an additional so far unidentified oxidising agent or some component which reduces the corrosion protection afforded by the build up of a corrosion product layer. Acquisition of improved electrochemical kinetic data for the mathematical model is now complete, and the model has been run for temperatures of 25 and 90 0 C, where it predicts steady corrosion rates of 19.3 and 180 μm/yr. The model has shown that the rate of attack is not influenced greatly by the depth of sediment, and that the component of corrosion caused by radiation is of the order of 7 mm over 1000 years. (author)

  5. Energy storage

    CERN Document Server

    Brunet, Yves

    2013-01-01

    Energy storage examines different applications such as electric power generation, transmission and distribution systems, pulsed systems, transportation, buildings and mobile applications. For each of these applications, proper energy storage technologies are foreseen, with their advantages, disadvantages and limits. As electricity cannot be stored cheaply in large quantities, energy has to be stored in another form (chemical, thermal, electromagnetic, mechanical) and then converted back into electric power and/or energy using conversion systems. Most of the storage technologies are examined: b

  6. Thermo-mechanical FE-analysis of butt-welding of a Cu-Fe canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Josefson, B.L.; Karlsson, L.; Lindgren, L.E.; Jonsson, M.

    1992-10-01

    In the Swedish nuclear waste program it has been proposed that spent nuclear fuel shall be placed in composite copper-steel canisters. These canisters will be placed in holes in tunnels located some 500 m underground in a rock storage. The canisters consists of two cylinders of 4850 mm length, one inner cylinder made of steel and one outer cylinder made of copper. The outer diameter of the canister is 880 mm and the wall thickness for each cylinder is 50 mm. At the storage, the steel cylinder, which contains the spent nuclear fuel, is placed inside the copper cylinder. Thereafter, a copper end is butt welded to the copper cylinder using electron beam welding. To obtain penetration through the thickness with this weld method a backing ring is placed at the inside of the copper cylinder. In the present paper, the temperature, strain and stress fields present during welding and after cooling after welding are calculated numerically using the FE-code NIKE-2D. The aim is to use the results of the present calculations to estimate the risk for creep fracture during the subsequent design life. A large strain formulation is employed for the calculation of transient and residual stresses in the canister based on the calculated history of the temperature field present in the canister during the welding process. The contact algorithm available in NIKE-2D is used to detect possible contact between the steel and copper cylinders during the welding. To simplify the numerical calculations and reduce the computational time, rotational symmetry is assumed. For large gap distances between the steel and copper cylinders the residual stress field is calculated to have a shape similar to that observed in butt welded pipes with maximum axial stress values at the yield stress level. For small gap distances the backing ring will come in contact with the steel cylinder which leads to large residual stresses in the backing ring. The maximum accumulated plastic strain in the weld metal and

  7. Numerical modelling and experimental studies of thermal behaviour of building integrated thermal energy storage unit in a form of a ceiling panel

    International Nuclear Information System (INIS)

    Jaworski, Maciej; Łapka, Piotr; Furmański, Piotr

    2014-01-01

    Highlights: • A new concept of heat storage in ventilation ducts is described. • Ceiling panel as a part of ventilation system is made of a composite with PCM. • A set-up for experimental investigation of heat storage unit was built. • Numerical model of heat transfer in the storage unit was developed. • Numerical code was validated on the base of experimental measurements. - Abstract: Objective: The paper presents a new concept of building integrated thermal energy storage unit and novel mathematical and numerical models of its operation. This building element is made of gypsum based composite with microencapsulated PCM. The proposed heat storage unit has a form of a ceiling panel with internal channels and is, by assumption, incorporated in a ventilation system. Its task is to reduce daily variations of ambient air temperature through the absorption (and subsequent release) of heat in PCM, without additional consumption of energy. Methods: The operation of the ceiling panel was investigated experimentally on a special set-up equipped with temperature sensors, air flow meter and air temperature control system. Mathematical and numerical models of heat transfer and fluid flow in the panel account for air flow in the panel as well as real thermal properties of the PCM composite, i.e.: thermal conductivity variation with temperature and hysteresis of enthalpy vs. temperature curves for heating and cooling. Proposed novel numerical simulator consists of two strongly coupled sub models: the first one – 1D – which deals with air flowing through the U-shaped channel and the second one – 3D – which deals with heat transfer in the body of the panel. Results: Spatial and temporal air temperature variations, measured on the experimental set-up, were used to validate numerical model as well as to get knowledge of thermal performance of the panel operating in different conditions. Conclusion: Preliminary results of experimental tests confirmed the ability of

  8. Mechanical Integrity of Canisters Using a Fracture Mechanics Approach

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Tomofumi; Guoxiang Zhang; Lanru Jing [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Land and Water Resources Engineering

    2006-07-15

    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

  9. 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

  10. Design, production and initial state of the canister

    International Nuclear Information System (INIS)

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

    2010-12-01

    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

  11. Burst Test Qualification Analysis of DWPF Canister-Plug Weld

    International Nuclear Information System (INIS)

    Gupta, N.K.; Gong, Chung.

    1995-02-01

    The DWPF canister closure system uses resistance welding for sealing the canister nozzle and plug to ensure leak tightness. The welding group at SRTC is using the burst test to qualify this seal weld in lieu of the shear test in ASME B ampersand PV Code, Section IX, paragraph QW-196. The burst test is considered simpler and more appropriate than the shear test for this application. Although the geometry, loading and boundary conditions are quite different in the two tests, structural analyses show similarity in the failure mode of the shear test in paragraph QW-196 and the burst test on the DWPF canister nozzle Non-linear structural analyses are performed using finite element techniques to study the failure mode of the two tests. Actual test geometry and realistic stress strain data for the 304L stainless steel and the weld material are used in the analyses. The finite element models are loaded until failure strains are reached. The failure modes in both tests are shear at the failure points. Based on these observations, it is concluded that the use of a burst test in lieu of the shear test for qualifying the canister-plug weld is acceptable. The burst test analysis for the canister-plug also yields the burst pressures which compare favorably with the actual pressure found during burst tests. Thus, the analysis also provides an estimate of the safety margins in the design of these vessels

  12. Underground gas storage Uelsen: Findings from planning, building and commissioning the surface buildings and structures; Untertagegasspeicher (UGS) Uelsen: Erkenntnisse aus Planung, Bau und Inbetriebnahme der obertaegigen Anlagen

    Energy Technology Data Exchange (ETDEWEB)

    Focke, H.; Brueggmann, R.; Mende, F.; Steinkraus, D.; Wauer, R. [BEB Erdgas und Erdoel GmbH, Hannover (Germany)

    1998-12-31

    The article describes the concepts of the plants and equipment and the specific features of the underground storage at Uelsen. The underground storage will be purpose-built as an H-gas storage in a nearly depleted sandstone deposit. At a nominal deliverability of 250.000 cubic m/h (Vn) the storage at Uelsen has more potential for expansion. This potential was taken into account by designing appropriate pressure stages, capacities, performance characteristics and space. (orig.). [Deutsch] Die nachfolgende Veroeffentlichung stellt das anlagentechnische Grundkonzept und die spezifischen Besonderheiten des UGS Uelsen dar. Der im suedwestlichen Niedersachsen als H-Gasspeicher in einer nahezu ausgefoerderten Buntsandsteinlagerstaette eingerichtete UGS Uelsen wird in mehreren Ausbaustufen bedarfsgerecht fertiggestellt. Bei einer Nennentnahmekapazitaet von 450.000 m{sup 3}/h (Vn) und einer Nenninjektionsleistung von 250.000 m{sup 3}/h (Vn) weist der UGS Uelsen noch weiteres Potential fuer Erweiterungen auf. Dieses Ausbaupotential wurde bei der Planung und dem Bau der bestehenden Anlagen durch Festlegung entsprechender Druckstufen, Kapazitaeten, Leistungsgroessen und Platzanordnungen beruecksichtigt. (orig.)

  13. Electropolishing decontamination system for high-level waste canisters

    International Nuclear Information System (INIS)

    Larson, D.E.; Berger, D.N.; Allen, R.P.; Bryan, G.H.; Place, B.G.

    1988-10-01

    As part of a US Department of Energy (DOE) project agreement with the Federal Ministry for Research and Technology (BMFT) in the Federal Republic of Germany (FRG). The Nuclear Waste Treatment Program at the Pacific Northwest Laboratory (PNL) is preparing 30 radioactive canisters containing borosilicate glass for use in high-level waste repository related tests at the Asse Salt Mine. After filling, the canisters will be welded closed and decontaminated in preparation for shipping to the FRG. Electropolishing was selected as the primary decontamination approach, and an electropolishing system with associated canister inspection equipment has been designed and fabricated for installation in a large hot cell. This remote electropolishing system, which is currently undergoing preliminary testing, is described in this report. 3 refs., 3 figs., 1 tab

  14. Drop tests of the Three Mile Island knockout canister

    International Nuclear Information System (INIS)

    Box, W.D.; Aaron, W.S.; Shappert, L.B.; Childress, P.C.; Quinn, G.J.; Smith, J.V.

    1987-01-01

    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 confirmed 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. This report presents the data generated and the results obtained from a series of four drop tests that included two drops with the test assembly in the vertical position and two drops with the assembly in the horizontal position

  15. A welding system for spent fuel canister lid

    International Nuclear Information System (INIS)

    Suikki, M.; Wendelin, T.

    2008-06-01

    The report presents a proposed welding system for spent fuel canister lids. The system is used for welding the copper lid to the copper overpack. The apparatus will be installed in the encapsulation plant. The report presents basic requirements for and implementation of the welding system, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components is not included. The report also describes actions for possible malfunction and fault conditions. Closing of the copper cylinder's lid is carried out by electron beam welding, which must be performed in vacuum. The welding system for spent fuel canister lid consists of two welding chambers, a canister docking system, an EB-welding machine with its accessories, a vacuum apparatus, as well as necessary auxiliary equipment. The system's equipment is housed in a welding room, an auxiliary system room, an operation control room, as well as mounted on the ceiling of a transfer corridor. One of the welding chambers is intended for carrying out test welding procedures and for calibration of welding parameters. The actual spent fuel canister lid welding chamber has a weldingready canister docked thereto in an airtight manner. The chamber is pumped for a vacuum, followed by closing the canister's copper lid and carrying out the lid welding process. The lid is brought into the chamber prior to docking the canister by means of a canister transfer trolley lifting gear. Lifting of the canister and rotating it during a welding process are also handled by means of the transfer trolley. The lid welding chamber houses equipment for the alignment and installation of the lid, as well as heating means for the top side of a copper overpack for ensuring a sufficient installation clearance between the lid and the overpack. The equipment not needed in the immediate vicinity of welding chambers, is

  16. Development of cold sprayed Cu coating for canister

    International Nuclear Information System (INIS)

    Kim, Hyung Jun; Kang, Yoon Ha

    2010-01-01

    Cold sprayed Cu deposition was studied for the application of outer part of canister for high level nuclear waste. Five commercially available pure Cu powders were analyzed and sprayed by high pressure cold spray system. Electrochemical corrosion test using potentiostat in 3.5% NaCl solution was conducted as well as microstructural analysis including hardness and oxygen content measurements. Overall evaluation of corrosion performance of cold sprayed Cu deposition is inferior to forged and extruded Cu plates, but some of Cu depositions are comparable to Cu plates. The simulated corrosion test in 200m underground cave is still in progress. The effect of cold spray process parameters was also studied and the results show that the type of nozzle is the most important other than powder feed rate, spray distance, and scan speed. 1/10 scale miniature of canister was manufactured confirming that the production of full scale canister is possible

  17. Prototype spent-fuel canister design, analysis, and test

    International Nuclear Information System (INIS)

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

  18. Test plan for the Sample Transfer Canister system

    International Nuclear Information System (INIS)

    Flanagan, B.D.

    1998-01-01

    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. These samples will be packaged and shipped in accordance with the US Department of Transportation (DOT) regulation 49 CFR 173.4, ''Exceptions for small quantities.'' The Sample Transfer Canister is of a ''French Can'' design, intended to be mated with a glove box for loading/unloading. Transport will typically take place north of the Wye Barricade between WRAP and the 222-S Laboratory. The Sample Transfer Canister will be shipped in an insulated ice chest, but the ice chest will not be a part of the small quantity package during prototype testing

  19. Waste canister closure welding using the inertia friction welding process

    International Nuclear Information System (INIS)

    Klein, R.F.; Siemens, D.H.; Kuruzar, D.L.

    1986-02-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it in a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive seal weld, the properties and thickness of which are at least equal to those of the canister material. This paper describes the inertia friction welding process and a proposed equipment design concept that will provide a positive, reliable, inspectable, and full thickness seal weld while providing easily maintainable equipment, even though the weld is made in a highly contaminated hot cell. All studies and tests performed have shown the concept to be highly feasible. 2 refs., 6 figs

  20. Combined storage system for LWR spent fuel and high-level waste

    International Nuclear Information System (INIS)

    Baxter, B.J.; Ganley, J.T.; Washington, J.A.

    1983-01-01

    The MODREX storage system consists of four basic elements: (1) the storage canister, (2) the storage module, (3) the storage cask, and (4) the transport cask. The storage canister is the heart of the system and, when used in combination with the module or either of the casks, allows the MODREX system to respond quickly to varied storage system requirements. The MODREX system can be used to hold either spent fuel assemblies or canistered solidified HLW. The ability to combine a basic storage canister with either a concrete module or a metal cask provides flexibility to meet a wide range of storage requirements. The spent fuel is stored in a dry, inert atmosphere, which essentially eliminates corrosion or deterioration of the cladding during extended storage periods. The storage canister and concrete storage module provide additional barriers against radioactivity release, enhancing long-term safety. Heat dissipation is passive, eliminating the need for additional emergency cooling systems or special redundancy. Modular, expandable construction permits minimum initial investment and capital carrying charges; additional capacity is built and paid for only as it is needed, retaining flexibility. 6 references, 2 figures, 1 table

  1. Construction and operation of a support facilities (Building 729) for operation/testing of a prototype accelerator/storage ring (XLS) and machine shop for the National Synchrotron Light Source at Brookhaven National Laboratory, Upton, New York

    International Nuclear Information System (INIS)

    1992-06-01

    Proposed action is to construct at BNL a 5,600-ft 2 support building, install and operate a prototypic 200 MeV accelerator and a prototypic 700 MeV storage ring within, and to construct and operate a 15 kV substation to power the building. The accelerator and storage ring would comprise the x-ray lithography source or XLS

  2. Development of Copper Canister through Cold Sprayed Coating Method

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

    2007-12-15

    General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future.

  3. Development of Copper Canister through Cold Sprayed Coating Method

    International Nuclear Information System (INIS)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

    2007-12-01

    General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future

  4. Chemical stability of copper-canisters in deep repository

    International Nuclear Information System (INIS)

    Ahonen, L.

    1995-12-01

    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.)

  5. Mechanical Integrity of Copper Canister Lid and Cylinder. Sensitivity study

    International Nuclear Information System (INIS)

    Karlsson, Marianne

    2002-08-01

    This report is part of a study of the mechanical integrity of canisters used for disposal of nuclear fuel waste. The overall objective is to determine and ensure the static and long-term strength of the copper canister lid and cylinder casing. The canisters used for disposal nuclear fuel waste of type BWR consists of an inner part (insert) of ductile cast iron and an outer part of copper. The copper canister is to provide a sealed barrier between the contents of the canister and the surroundings. The study in this report complements the finite element analyses performed in an earlier study. The analyses aim to evaluate the sensitivity of the canister to tolerances regarding the gap between the copper cylinder and the cast iron insert. Since great uncertainties regarding the material's long term creep properties prevail, analyses are also performed to evaluate the effect of different creep data on the resulting strain and stress state. The report analyses the mechanical response of the lid and flange of the copper canister when subjected to loads caused by pressure from swelling bentonite and from groundwater at a depth of 500 meter. The loads acting on the canister are somewhat uncertain and the cases investigated in this report are possible cases. Load cases analysed are: Pressure 15 MPa uniformly distributed on lid and 5 MPa uniformly distributed on cylinder; Pressure 5 MPa uniformly distributed on lid and 15 MPa uniformly distributed on cylinder; Pressure 20 MPa uniformly distributed on lid and cylinder; and Side pressures 10 MPa and 20 MPa uniformly distributed on part of the cylinder. Creep analyses are performed for two of the load cases. For all considered designs high principal stresses appear on the outside of the copper cylinder in the region from the weld down to the level of the lid lower edge. Altering the gap between lid and cylinder and/or between cylinder and insert only marginally affects the resulting stress state. Fitting the lid in the cylinder

  6. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    International Nuclear Information System (INIS)

    1990-03-01

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

  7. Multi Canister Overpack (MCO) Closure Welding Process Parameter Development and Qualification

    International Nuclear Information System (INIS)

    CANNELL, G.R.

    2003-01-01

    One of the Department of Energy's (DOE) top priorities at the Hanford Site (southeastern Washington state), is the processing of more than 2,000 tons of spent nuclear fuel (SNF) into large stainless steel containers called Multi-Canister Overpacks (MCO). Packaging into MCO's will assist in the safe and economic disposition of SNF and greatly reduce risk to the environment. Packaged fuel will be removed from close proximity to the Columbia River to a more suitable area of the site where it will be stored on an interim basis. Eventually, the fuel will be transferred to the federal geologic repository for long-term storage. 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 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 describes the effort to develop and qualify the root-pass technique associated with the MCO final closure weld

  8. 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)

    Boergesson, L.

    1992-07-01

    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

  9. Underground gas storage Uelsen: Findings from planning, building and commissioning. Part 1: Deposit; Untertagegasspeicher Uelsen: Erkenntnisse aus Planung, Bau und Inbetriebnahme. Teil 1: Lagerstaette

    Energy Technology Data Exchange (ETDEWEB)

    Wallbrecht, J.; Beckmann, H.; Reiser, H.; Wilhelm, R. [BEB Erdgas und Erdoel GmbH, Hannover (Germany)

    1998-12-31

    The underground gas storage at Uelsen which was built as a H-gas storage in a former variegated sandstone gasfield in Western Lower Saxony close to the town of Nordhorn has added to the gas supply system of the BEB Erdgas and Erdoel GmbH. The underground storage is connected to the Bunde-Rheine transport pipeline BEB-grid gas system by a 27 km pipeline and is a consequent expansion of BEB`s underground storage/transport system. Planning, building and commissioning were handled by BEB. Findings to date are described. [Deutsch] Der Untertagegasspeicher (UGS) Uelsen, der in einem ehemaligen Buntsandstein Gasfeld im westlichen Niedersachsen in der Naehe der Stadt Nordhorn als H-Gasspeicher eingerichtet wurde, hat die BEB Erdgas und Erdoel GmbH eine weitere Staerkung ihres Gasversorgungssystems erreicht. Der UGS Uelsen ist ueber eine 27 km lange Anbindungsleitung mit der zum BEB - Ferngasleitungssystems gehoerenden Bunde-Rheine Transportleitung verbunden und stellt eine konsequente Erweiterung des BEB Untertagegasspeicher-/Transportsystems dar. Planung, Bau und Inbetriebnahme erfolgten durch BEB im Rahmen einer integrierten bereichsuebergreifenden Projektbearbeitung. Die hierbei gewonnenen Erkenntnisse werden im Folgenden fuer den Untertagebereich dargestellt. (orig.)

  10. Quantifying demand flexibility of power-to-heat and thermal energy storage in the control of building heating systems

    NARCIS (Netherlands)

    Finck, C.J.; Li, R.; Kramer, R.P.; Zeiler, W.

    2018-01-01

    In the future due to continued integration of renewable energy sources, demand-side flexibility would be required for managing power grids. Building energy systems will serve as one possible source of energy flexibility. The degree of flexibility provided by building energy systems is highly

  11. Effect of Heat and Electricity Storage and Reliability on Microgrid Viability:A Study of Commercial Buildings in California and New York States

    Energy Technology Data Exchange (ETDEWEB)

    Stadler, Michael; Marnay, Chris; Siddiqui, Afzal; Lai, Judy; Coffey, Brian; Aki, Hirohisa

    2008-12-01

    In past work, Berkeley Lab has developed the Distributed Energy Resources Customer Adoption Model (DER-CAM). Given end-use energy details for a facility, a description of its economic environment and a menu of available equipment, DER-CAM finds the optimal investment portfolio and its operating schedule which together minimize the cost of meeting site service, e.g., cooling, heating, requirements. Past studies have considered combined heat and power (CHP) technologies. Methods and software have been developed to solve this problem, finding optimal solutions which take simultaneity into account. This project aims to extend on those prior capabilities in two key dimensions. In this research storage technologies have been added as well as power quality and reliability (PQR) features that provide the ability to value the additional indirect reliability benefit derived from Consortium for Electricity Reliability Technology Solutions (CERTS) Microgrid capability. This project is intended to determine how attractive on-site generation becomes to a medium-sized commercial site if economical storage (both electrical and thermal), CHP opportunities, and PQR benefits are provided in addition to avoiding electricity purchases. On-site electrical storage, generators, and the ability to seamlessly connect and disconnect from utility service would provide the facility with ride-through capability for minor grid disturbances. Three building types in both California and New York are assumed to have a share of their sensitive electrical load separable. Providing enhanced service to this load fraction has an unknown value to the facility, which is estimated analytically. In summary, this project began with 3 major goals: (1) to conduct detailed analysis to find the optimal equipment combination for microgrids at a few promising commercial building hosts in the two favorable markets of California and New York; (2) to extend the analysis capability of DER-CAM to include both heat and

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

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2006-12-01

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

  13. A study on the thermal design of the advanced multi purpose canister

    International Nuclear Information System (INIS)

    Jun, Eun Ju; Chang, S. H.

    2004-01-01

    The storage, transportation and disposal of pressurized water reactor (PWR) spent fuel is an important issue in nuclear industry for safe and extended nuclear power plant operation. However, since wet-storage facilities almost reach to their capacity, dry-storage method, especially using a Multi Purpose Canister (MPC), has been highlighted because it gives better economic and safety point of view. In this study, an advanced design of MPC was proposed through an increase in storage capacity from 26 to 37 PWR fuel assemblies per MPC. Thermal analysis was also carried out by FLUENT 6 code to confirm the temperature criteria of the MPC. From the results, conceptual design of 37 PWR fuel assemblies can reduce the mass per fuel assembly ratio about 19 % compared to the 26 assemblies, and maximum temperature developed at the center of fuel assemblies shows within the range of design criteria. Based on the analysis, it is considered that the advanced MPC design well conforms to the safety requirement and it gives rise to economical gain because of reduction in mass per fuel assembly ratio

  14. Effects of brine migration on waste storage systems. Final report

    International Nuclear Information System (INIS)

    Gaffney, E.S.; Nickell, R.E.

    1979-01-01

    Processes which can lead to mobilization of brine adjacent to spent fuel or nuclear waste canisters and some of the thermomechanical consequences have been investigated. Velocities as high as 4 x 10 -7 m s -1 (13 m y -1 ) are calculated at the salt/canister boundary. As much as 40 liters of pure NaCl brine could accumulate around each canister during a 10-year storage period. Accumulations of bittern brines would probably be less, in the range of 2 to 5 liters. With 0.5% water, NaCl brine accumulation over a 10-year storage cycle around a spent fuel canister producing 0.6 kW of heat is expected to be less than 1 liter for centimeter-size inclusions and less than 0.5 liter for millimeter-size inclusions. For bittern brines, about 25 years would be required to accumulate 0.4 liter. The most serious mechanical consequence of brine migration would be the increased mobility of the waste canister due to pressure solution. In pressure solution enhanced deformation, the existence of a thin film of fluid either between grains or between media (such as between a canister and the salt) provides a pathway by which the salt can be redistributed leading to a marked increase in strain rates in wet rock relative to dry rock. In salt, intergranular water will probably form discontinuous layers rather than films so that they would dominate pressure solution. A mathematical model of pressure solution indicates that pressure solution will not lead to appreciable canister motions except possibly in fine grained rocks (less than 10 -4 m). In fine grained salts, details of the contact surface between the canister and the salt bed may lead to large pressure solution motions. A numerical model indicates that heat transfer in the brine layer surrounding a spent fuel canister is not conduction dominated but has a significant convective component

  15. 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.

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

    International Nuclear Information System (INIS)

    1995-01-01

    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

  17. Examination of sludge from the Hanford K Basins fuel canisters

    International Nuclear Information System (INIS)

    Makenas, B.J.

    1998-01-01

    Samples of sludges with a high uranium content have been retrieved from the fuel canisters in the Hanford K West and K East basins. The composition of these samples contrasts markedly with the previously reported content of sludge samples taken from the K East basin floor. Chemical composition, chemical reactivity, and particle size of sludge are summarized in this paper

  18. Effects of glacial meltwater on corrosion of copper canisters

    International Nuclear Information System (INIS)

    Ahonen, L.; Vieno, T.

    1994-08-01

    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.)

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

    International Nuclear Information System (INIS)

    1983-04-01

    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.)

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

    International Nuclear Information System (INIS)

    Sjoeblom, R.; Hermansson, H.P.; Amcoff, Oe.

    1995-01-01

    In the Swedish waste management programme, the copper canister is expected to provide containment of the radionuclides for a very long time, perhaps million 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 layers 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. 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. 20 refs, 3 figs, 1 tab

  1. Retrievable storage concept designs. Final report

    International Nuclear Information System (INIS)

    Nickell, R.E.

    1979-01-01

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

  2. Transportation system benefits of early deployment of a 75-ton multipurpose canister system

    International Nuclear Information System (INIS)

    Wankerl, M.W.; Schmid, S.P.

    1995-01-01

    In 1993 the US Civilian Radioactive Waste Management System (CRWMS) began developing two multipurpose canister (MPC) systems to provide a standardized method for interim storage and transportation of spent nuclear fuel (SNF) at commercial nuclear power plants. One is a 75-ton concept with an estimated payload of about 6 metric tons (t) of SNF, and the other is a 125-ton concept with an estimated payload of nearly 11 t of SNF. These payloads are two to three times the payloads of the largest currently certified US rail transport casks, the IF-300. Although is it recognized that a fully developed 125-ton MPC system is likely to provide a greater cost benefit, and radiation exposure benefit than the lower-capacity 75-ton MPC, the authors of this paper suggest that development and deployment of the 75-ton MPC prior to developing and deploying a 125-ton MPC is a desirable strategy. Reasons that support this are discussed in this paper

  3. The development of a Martian atmospheric Sample collection canister

    Science.gov (United States)

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

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

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

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    2000-06-01

    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

  5. Experimental and numerical investigations of a zeolite 13X/water reactor for solar heat storage in buildings

    International Nuclear Information System (INIS)

    Tatsidjodoung, Parfait; Le Pierrès, Nolwenn; Heintz, Julien; Lagre, Davy; Luo, Lingai; Durier, François

    2016-01-01

    Highlights: • An open cycle heat storage using zeolite 13X/H 2 O is investigated. • A 1D reactor model is developed and compared to experimental results. • 40 kg batches generated up to 38 °C of temperature lift during 8 h of discharging. - Abstract: This paper addresses the thermal performances of a zeolite-based open sorption heat storage system to provide thermal energy for space heating needs. The study focuses on the experimentation of a significant scale prototype using zeolite 13X/H 2 O as the reactive pair, and on the development of a 1D mathematical model used to predict both the charging (desorption) and the discharging (adsorption) processes occurring inside the storage unit. The experimental campaigns and the numerical results lead to some promising conclusions on the thermal performances of such a storage unit. With 40 kg of zeolite, a temperature lift of 38 °C on average at the outlet of each zeolite’s vessel during 8 h was achieved during the discharging with an airflow inlet at 20 °C, 10 g/kg of dry air of specific humidity and a flow rate of 180 m 3 /h. Some discrepancies between the experimental and simulation results were observed during both the charging and discharging tests, and were explained.

  6. Low-Cost Bio-Based Phase Change Materials as an Energy Storage Medium in Building Envelopes

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, Kaushik [ORNL; Abhari, Mr. Ramin [Renewable Energy Group, Inc.; Shukla, Dr. Nitin [Fraunhofer USA, Center for Sustainable Energy Systems (CSE), Boston; Kosny, Dr. Jan [Fraunhofer USA, Center for Sustainable Energy Systems (CSE), Boston

    2015-01-01

    A promising approach to increasing the energy efficiency of buildings is the implementation of phase change material (PCM) in building envelope systems. Several studies have reported the energy saving potential of PCM in building envelopes. However, wide application of PCMs in building applications has been inhibited, in part, by their high cost. This article describes a novel paraffin product made of naturally occurring fatty acids/glycerides trapped into high density polyethylene (HDPE) pellets and its performance in a building envelope application, with the ultimate goal of commercializing a low-cost PCM platform. The low-cost PCM pellets were mixed with cellulose insulation, installed in external walls and field-tested under natural weatherization conditions for a period of several months. In addition, several PCM samples and PCM-cellulose samples were prepared under controlled conditions for laboratory-scale testing. The laboratory tests were performed to determine the phase change properties of PCM-enhanced cellulose insulation both at microscopic and macroscopic levels. This article presents the data and analysis from the exterior test wall and the laboratory-scale test data. PCM behavior is influenced by the weather and interior conditions, PCM phase change temperature and PCM distribution within the wall cavity, among other factors. Under optimal conditions, the field data showed up to 20% reduction in weekly heat transfer through an external wall due to the PCM compared to cellulose-only insulation.

  7. Form-Stable Phase Change Materials Based on Eutectic Mixture of Tetradecanol and Fatty Acids for Building Energy Storage: Preparation and Performance Analysis

    Directory of Open Access Journals (Sweden)

    Yiran li

    2013-10-01

    Full Text Available This paper is focused on preparation and performance analysis of a series of form-stable phase change materials (FSPCMs, based on eutectic mixtures as phase change materials (PCMs for thermal energy storage and high-density polyethylene (HDPE-ethylene-vinyl acetate (EVA polymer as supporting materials. The PCMs were eutectic mixtures of tetradecanol (TD–c