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Sample records for canister storage building

  1. Canister storage building natural phenomena design loads

    International Nuclear Information System (INIS)

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

  2. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

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

  3. Canister storage building hazard analysis report

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  4. Canister storage building trade study. Final report

    International Nuclear Information System (INIS)

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

  5. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-23

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

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

    International Nuclear Information System (INIS)

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

  9. 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.

  10. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

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

  11. Canister Storage Building (CSB) Hazard Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    2000-03-16

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

  12. Canister Storage Building (CSB) Hazard Analysis Report

    International Nuclear Information System (INIS)

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

  13. 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.

  14. Spent nuclear fuel canister storage building conceptual design report

    International Nuclear Information System (INIS)

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

  15. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    HOLLENBECK, R.G.

    2000-05-08

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    GARRISON, R.C.

    2000-11-28

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-17

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

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

    CERN Document Server

    Lewis, M E

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  13. Airborne Effluent Monitoring System Certification for New Canister Storage Building Ventilation Exhaust Stack

    International Nuclear Information System (INIS)

    Pacific Northwest National Laboratory conducted three of the six tests needed to verify that the effluent monitoring system for the new Canister Storage Building ventilation exhaust stack meets applicable regulatory performance criteria for air sampling systems at nuclear facilities. These performance criteria address both the suitability of the location for the air-sampling probe and the transport of the sample to the collection devices. The criteria covering the location for the air-sampling probe ensure that the contaminants in the stack are well mixed with the airflow at the probe location such that the extracted sample represents the whole. The sample-transport criteria ensure that the sampled contaminants are quantitatively delivered to the collection device. The specific performance criteria are described in detail in this report. The tests reported here cover the contaminant tracer uniformity and particle delivery performance criteria. These criteria were successfully met. The other three tests were conducted by the start-up staff of Duke Engineering and Services Hanford Inc. (DESH) and reported elsewhere. The Canister Storage Building is located in the 200 East Area of the U.S. Department of Energy's Hanford Site near Richland, Washington. The new air-exhaust system was built under the W379 Project. The air sampling system features a probe with a single shrouded sampling nozzle, a sample delivery line, and a filter holder to collect the sample

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    GARRISON, R.C.

    2000-09-22

    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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  1. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Criticality safety calculations of storage canisters

    International Nuclear Information System (INIS)

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

  6. Temperature history for canistered fuel lag storage areas during the loss of cooling air at the receiving and handling building of the MRS Facility

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory has analyzed the temperature history at a canistered fuel lag storage area during a postulated failure of the cooling system. A two-dimensional analysis was performed using the GE2D computer code, which accounts for thermal radiation, conduction, and convective heat transfer. The results indicate that the system may not reach a steady-state condition because heat transfer through the top and the bottom of the system is not enough to remove the energy generated in the canistered fuel. Although limits for abnormal operation have not been set, the temperatures do not reach limiting conditions for normal operation for 32 hours, which should be enough time to repair the cooling system

  7. Structural Sensitivity of Dry Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-27

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wittman, Richard S.

    2013-09-20

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

  9. 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.

  10. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Canister Storage Building (CSB) Technical Safety Requirements

    International Nuclear Information System (INIS)

    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

  15. 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.

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Kerslake, Thomas W.; Ibrahim, Mounir B.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

  19. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Use of ground penetrating radar for the design verification of 'as-built' spent fuel storage canisters in Argentina

    International Nuclear Information System (INIS)

    This paper documents the results of a demonstration exercise to enable the IAEA to assess the use of Ground Penetrating Radar (GPR) for Safeguards requirements, such as design verification. This has been provided in the form of a demonstration of the equipment and the technique in operation at two sites in Argentina, the Cemetary at Ezeiza Centro Atomico and the spent-fuel (SF) dry-storage canisters at Embalse NPS, In April 1995. The work subsequently lead to the technology being employed in a formal design information verification exercise of 42 canisters at the Embalse facility in December 1995. The demonstration and subsequent verification work have shown that complete surveys around a building, a deck or over the entire area of a canister are practical and not excessively time-consuming. The requirement for lower frequency probes for maximum penetration when interrogating soils has also been explained. It is concluded that ground radar has provided very useful evidence on the Canisters at Embalse. It is also concluded that, given these demonstrations, their results, and those of the subsequent formal inspection, there are probably a number of situations where Ground Penetrating Radar could be employed effectively by the Agency in its Safeguards role. (author)

  3. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  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)

    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)

    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. Shielded canister transporter equipment acceptance test operations

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

  16. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Eddy Current for Sizing Cracks in Canisters for Dry Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Jones, Anthony M.; Pardini, Allan F.

    2014-01-01

    The storage of used nuclear fuel (UNF) in dry canister storage systems (DCSSs) at Independent Spent Fuel Storage Installations (ISFSI) sites is a temporary measure to accommodate UNF inventory until it can be reprocessed or transferred to a repository for permanent disposal. Policy uncertainty surrounding the long-term management of UNF indicates that DCSSs will need to store UNF for much longer periods than originally envisioned. Meanwhile, the structural and leak-tight integrity of DCSSs must not be compromised. The eddy current technique is presented as a potential tool for inspecting the outer surfaces of DCSS canisters for degradation, particularly atmospheric stress corrosion cracking (SCC). Results are presented that demonstrate that eddy current can detect flaws that cannot be detected reliably using standard visual techniques. In addition, simulations are performed to explore the best parameters of a pancake coil probe for sizing of SCC flaws in DCSS canisters and to identify features in frequency sweep curves that may potentially be useful for facilitating accurate depth sizing of atmospheric SCC flaws from eddy current measurements.

  20. The Effect of Weld Residual Stress on Life of Used Nuclear Fuel Dry Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Ronald G. Ballinger; Sara E. Ferry; Bradley P. Black; Sebastien P. Teysseyre

    2013-08-01

    With the elimination of Yucca Mountain as the long-term storage facility for spent nuclear fuel in the United States, a number of other storage options are being explored. Currently, used fuel is stored in dry-storage cask systems constructed of steel and concrete. It is likely that used fuel will continue to be stored at existing open-air storage sites for up to 100 years. This raises the possibility that the storage casks will be exposed to a salt-containing environment for the duration of their time in interim storage. Austenitic stainless steels, which are used to construct the canisters, are susceptible to stress corrosion cracking (SCC) in chloride-containing environments if a continuous aqueous film can be maintained on the surface and the material is under stress. Because steel sensitization in the canister welds is typically avoided by avoiding post-weld heat treatments, high residual stresses are present in the welds. While the environment history will play a key role in establishing the chemical conditions for cracking, weld residual stresses will have a strong influence on both crack initiation and propagation. It is often assumed for modeling purposes that weld residual stresses are tensile, high and constant through the weld. However, due to the strong dependence of crack growth rate on stress, this assumption may be overly conservative. In particular, the residual stresses become negative (compressive) at certain points in the weld. The ultimate goal of this research project is to develop a probabilistic model with quantified uncertainties for SCC failure in the dry storage casks. In this paper, the results of a study of the residual stresses, and their postulated effects on SCC behavior, in actual canister welds are presented. Progress on the development of the model is reported.

  1. 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)

    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

  2. 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.

  3. CFD modeling of natural convection within dry spent nuclear fuel storage canisters

    International Nuclear Information System (INIS)

    One of the interim storage configurations being considered for aluminum-clad foreign research reactor fuel, such as the Material and Testing Reactor (MTR) design, is in a dry storage facility. To support design studies of storage options, a computational and experimental program was conducted at the Savannah River Site (SRS). The objective was to develop computational fluid dynamics (CFD) models which would be benchmarked using data obtained from a full scale heat transfer experiment conducted in the SRS Experimental Thermal Fluids Laboratory. The current work documents the CFD approach and presents comparison of results with experimental data. CFDS-FLOW3D (version 3.3) CFD code has been used to model the 3-dimensional convective velocity and temperature distributions within a single dry storage canister of MTR fuel elements. The analysis was made for the cases with q double-prime ' = 100 or 137 watts per MTR fuel element (equivalent to 25 or 35 kW/m3) using different convective boundary conditions around the canister wall and different cooling gases (N2 or He). For the present analysis, the Boussinesq approximation was used for the consideration of buoyancy-driven natural convection. Comparison of the CFD code can be used to predict reasonably accurate flow and thermal behavior of a typical foreign research reactor fuel stored in a dry storage facility

  4. Modellization of Metal Hydride Canister for Hydrogen Storage

    Directory of Open Access Journals (Sweden)

    Rocio Maceiras

    2015-06-01

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

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

    International Nuclear Information System (INIS)

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

  6. Stress corrosion cracking of stainless-steel canister for concrete cask storage of spent fuel

    Science.gov (United States)

    Tani, Jun-ichi; Mayuzumi, Masami; Hara, Nobuyoshi

    2008-09-01

    Resistance to external stress corrosion cracking (ESCC) and crevice corrosion were examined for various candidate canister materials in the spent fuel dry storage condition using concrete casks. A constant load ESCC test was conducted on the candidate materials in air after deposition of simulated sea salt particles on the specimen gage section. Highly corrosion resistant stainless steels (SS), S31260 and S31254, did not fail for more than 46 000 h at 353 K with relative humidity of 35%, although the normal stainless steel, S30403 SS failed within 500 h by ESCC. Crevice corrosion potentials of S31260 and S31254 SS became larger than 0.9 V (SCE) in synthetic sea water at temperatures below 298 K, while those of S30403 and S31603 SS were less than 0 V (SCE) at the same temperature range. No rust was found on S31260 and S31254 SS specimens at temperatures below 298 K in the atmospheric corrosion test, which is consistent with the temperature dependency of crevice corrosion potential. From the test result, the critical temperature of atmospheric corrosion was estimated to be 293 K for both S31260 and S31254 SS. Utilizing the ESCC test result and the critical temperature, together with the weather station data and the estimated canister wall temperature, the integrity of canister was assessed from the view point of ESCC.

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

    International Nuclear Information System (INIS)

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

  8. 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.

  9. 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

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

    International Nuclear Information System (INIS)

    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

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

    Directory of Open Access Journals (Sweden)

    Mohd Iswadi Ismail

    2014-04-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  15. Spent nuclear fuel storage building

    International Nuclear Information System (INIS)

    Outer earthquake proof walls of a building are formed in the running direction of a bridge travelling type crane. Inner earthquake proof walls are disposed in the direction perpendicular to the running direction of the crane. An area for constructing an air ventilation tower is formed in the running direction of the crane at the central portion of the building. The area surrounded by the area for constructing air ventilation tower and the earthquake proof inner walls is defined as a cask storage area. Such cask storage areas are disposed symmetrically on both sides of the air ventilation tower. The crane is disposed on the roof of the building, stands on rails which are laid on the outer walls of the building and runs over the air ventilation tower. Since the crane is disposed on the roof of the building, the inner earthquake proof walls of the building can be constructed in the inside of the building, and a high earthquake proofness can be provided to the building itself. (I.N.)

  16. 2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

    International Nuclear Information System (INIS)

    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

  17. Experimental and Computational Investigations of Phase Change Thermal Energy Storage Canisters

    Science.gov (United States)

    Ibrahim, Mounir; Kerslake, Thomas; Sokolov, Pavel; Tolbert, Carol

    1996-01-01

    Two sets of experimental data are examined in this paper, ground and space experiments, for cylindrical canisters with thermal energy storage applications. A 2-D computational model was developed for unsteady heat transfer (conduction and radiation) with phase-change. The radiation heat transfer employed a finite volume method. The following was found in this study: (1) Ground Experiments: the convection heat transfer is equally important to that of the radiation heat transfer; radiation heat transfer in the liquid is found to be more significant than that in the void; including the radiation heat transfer in the liquid resulted in lower temperatures (about 15 K) and increased the melting time (about 10 min.); generally, most of the heat flow takes place in the radial direction. (2) Space Experiments: radiation heat transfer in the void is found to be more significant than that in the liquid (exactly the opposite to the Ground Experiments); accordingly, the location and size of the void affects the performance considerably; including the radiation heat transfer in the void resulted in lower temperatures (about 40 K).

  18. Demonstrative drop tests of transport and storage full-scale canisters with high corrosion-resistant material

    International Nuclear Information System (INIS)

    The concrete modular dry storage technologies are becoming widely-used, aiming at better economic performances. In 1997, we commenced a research program of the demonstration test for interim storage of spent fuel, mainly involving concrete cask storage technologies, particularly aiming at the realization of dry storage away from reactor in 2010. Key issues of this research include safety standards in operation and maintenance during storage and unloading/loading for transportation, long-term integrity of metal canister and concrete materials, and so on. To propose safety standards for concrete cask structures, systems, components, the demonstration program for qualification of concrete cask performance under the normal, ab-normal and accidental events was successfully terminated. Especially, due to the lack of the experimental studies related to tipping-over or drop event scenarios, in this research program, the demonstration drop test program using double-lid welded multi-purpose canister (MPCs) was executed, with the aim of obtaining basic data for regulating safety. This paper introduces the summary of the CRIEPI's drop test program

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

    International Nuclear Information System (INIS)

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

  20. Study on commercial realization of concrete cask for interim storage of spent nuclear fuel. Proposal of examination methodology for stainless steel canister lid weldment

    International Nuclear Information System (INIS)

    To realize the early utilization of economical and module storage methodology, the Japan Society of Mechanical Engineers (JSME) had already developed and issued a Code for Construction of SNF Facilities - Rules on Concrete Cask - (JSME S FB1-2003) in 2003. On the other hand, the former competent authority (Nuclear and Industrial Safety Agency) issued the safety requirements for concrete cask design in 2004. For confinement requirements, Ultrasonic Test (UT) should be used to inspect the lid weldment in addition to Dye Penetrant Test (PT). To contribute to the JSME codification activities for adjustment between the safety requirements and the JSME design code, we performed mock-up canister weldment tests considering the vapor from the hot water in the spent fuel canister, the UT tests with canister lid weldment at high temperature and the structural analysis of the canister under hypothetical drop conditions. As a result, the draft amendment of the JSME design code was proposed. (author)

  1. The concrete canister program

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-07

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Development of measurement technology of chlorine attached on canister using laser. Remote measurement in narrow space for the application during storage of spent fuel

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) at the surface of a stainless steel canister is a current issue in the interim storage of spent fuel by a concrete cask. The concentration measurement of salt attached on a canister is one of the important methods for inspection of environmental condition for SCC occurrence. Laser-induced breakdown spectroscopy (LIBS) is an attractive method for the concentration measurement of salt attached on the candidate material of a canister. In order to adopt the LIBS to the actual equipment, the remote measurement is needed because the surface of the canister is under radioactive and hot environmental condition. In this study, the remote measurement by using open pass LIBS was performed, and the prototype of compact devices for remote LIBS has been developed. The developed device for laser focusing can be inserted in the narrow space simulating the space between the concrete body and the canister. The results of open pass LIBS showed that the chlorine emission spectrum was measured in the narrow space when the distance from laser device to the measurement points was over 20 m. (author)

  6. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

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

  7. Modeling void growth and movement with phase change in thermal energy storage canisters

    Science.gov (United States)

    Darling, Douglas; Namkoong, David; Skarda, J. R. L.

    1993-01-01

    A scheme was developed to model the thermal hydrodynamic behavior of thermal energy storage salts. The model included buoyancy, surface tension, viscosity, phases change with density difference, and void growth and movement. The energy, momentum, and continuity equations were solved using a finite volume formulation. The momentum equation was divided into two pieces. The void growth and void movement are modeled between the two pieces of the momentum equations. Results showed this scheme was able to predict the behavior of thermal energy storage salts.

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

    Science.gov (United States)

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

    2015-03-01

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

  10. 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)

    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

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

    International Nuclear Information System (INIS)

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

  12. Study on interim storage of spent nuclear fuel by concrete cask for practical use. Feasibility study on prevention of chloride induced stress corrosion cracking for type304L stainless steel canister

    International Nuclear Information System (INIS)

    For the practical use of the concrete cask storage method, remaining issues are preventive design (monitoring, inspection and countermeasures) and its demonstration of the Stress Corrosion Cracking (SCC) on the canister surface. Scenarios to maintain its confinement function of the canister made of the conventional SUS 304L materials during storage period were established by keeping the salt density on the canister surface not be exceed its critical salt density to initiate SCC or by controlling the crack propagation if the salt density exceeded the critical value. Furthermore the feasibility of the scenarios were demonstrated by tests defining the critical salt density for the SCC initiation and by tests of crack propagation based on metrological data of representative coastal sites in Japan. On top of that, methods of reduction of welding residual stress to prevent SCC were demonstrated by SCC tests using small scale test model made of SUS 304L simulating wall thickness of the real canister and welding methods. (author)

  13. Multi-canister overpack: additional NRC requirements

    International Nuclear Information System (INIS)

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

  14. 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

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

    International Nuclear Information System (INIS)

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

  16. Shippingport Spent Fuel Canister System Description

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-03-27

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

  17. Shippingport Spent Fuel Canister System Description

    International Nuclear Information System (INIS)

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

  18. Development of measurement technology of chlorine attached on canister using laser. Development of measurement device and quantification of attached chlorine concentration for measurement scenario during storage of spent fuel

    International Nuclear Information System (INIS)

    Chlorine concentration of salt attached on stainless steel, which is a candidate for the canister material of concrete cask, was measured by laser-induced breakdown spectroscopy (LIBS) for the purpose of the measurement during the storage of spent fuel. In order to measure a salt in a narrow gap between a canister and a concrete body, the compact optical device was developed to measure emission spectra by adopting the optical setup of a collinear configuration. In addition, LIBS system was developed for detecting chlorine emission spectra. Using a compact optical device, salt attached on a stainless-steel was measured vertically in the narrow space simulating the gap between a canister and a concrete body. The results of LIBS were coincident with the result of ion chromatography. These results show the possibility of quantitative measurement of chlorine on canister by LIBS for monitoring the environmental condition for stress corrosion cracking of canister in the concrete body. The series of LIBS experiments for the purpose of the measurement of salt attached on the canister made clear the agenda, such as the acquisition of the data set of calibration curve in various experimental conditions for the qualification of LIBS results, for adopting the method to an concrete cask. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Baker, R.B.

    1997-04-30

    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.

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

    International Nuclear Information System (INIS)

    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

  1. Thermal calculations for the study of the heat evacuation in the vaults building of the centralised temporary storage (ATC)

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.

    1999-08-12

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-14

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

  6. Multi-Canister Overpack (MCO) Topical Report

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Mohd Iswadi Ismail; Khairul Osman; Siti Fatimah Ibrahim; Farah Hanan Fatihah Jaafar; Nur Azianie Abd Ghani; Fazly Ann Zainalabidin; Abas Mazni Othman

    2014-01-01

    The objective of this study was to evaluate the potential of using single tubing-type canister on sperm quality. Semen was collected from the Bali cattle bull by electroejaculation technique and was cryopreserved in liquid nitrogen using slow freezing cryopreservation method. Two type of canister volume was used in this study; commercial canister (342.25π x 278 mm²) and single tubing-type canister (4π x 90 mm²). Makler counting chamber and computer assisted sperm analyzer (CASA) were used to ...

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

    International Nuclear Information System (INIS)

    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

  11. 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...

  12. Analysis of K west basin canister gas

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

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

  13. Ice slurry based thermal storage in multifunctional buildings

    Science.gov (United States)

    Wang, M. J.; Kusumoto, N.

    Ice slurry based thermal storage plays an important role in reshaping patterns of electricity use for space cooling and heating. It offers inherent advantages in energy efficiency, operating savings, load follow-up and flexible installation over conventional thermal storage technologies. This paper provides discussions on the generation mechanism and performance of ice slurry, as well as the operation principle of the ice slurry based thermal storage system. Details of the system design, control strategy and operation performance are given through a case study on a recent installation in Herbis Osaka, the largest simple building complex in Japan. An evaluation of different installations with ice slurry thermal storage reveals that it is a rewarding technology that provides significant operating savings for the building air-conditioning and improves energy utilization efficiency in modern society.

  14. Measurement of Moisture Storage Parameters of Building Materials

    OpenAIRE

    M. Jiřičková; Černý, R.; P. Rovnaníková

    2003-01-01

    The moisture storage parameters of three different building materials: calcium silicate, ceramic brick and autoclaved aerated concrete, are determined in the hygroscopic range and overhygroscopic range. Measured sorption isotherms and moisture retention curves are then combined into moisture storage functions using the Kelvin equation. A comparison of measured results with global characteristics of the pore space obtained by mercury intrusion porosimetry shows a reasonable agreement; the medi...

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

    DEFF Research Database (Denmark)

    Christensen, Jørgen Erik; Janssen, Hans

    2010-01-01

    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...... 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...... purposes. Reduction of dehumidification load: In an effort to reduce the necessary dehumidification, a number of thermal measures are investigated first. This primarily focuses on the influences of additional insulation in walls, roof and floor. Overall, the effects of extra insulation on the average...

  16. Building a mass storage system for physics applications

    International Nuclear Information System (INIS)

    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

  17. Passive hygrothermal control of a museum storage building

    DEFF Research Database (Denmark)

    Christensen, Jørgen Erik; Janssen, Hans

    2011-01-01

    For optimal conservation of the stored objects, museum storage buildings require a very stable interior climate, with only minimal and slow variations in temperature and relative humidity. Often extensive HVAC is installed to provide such stable indoor conditions, which results in a great amout of...

  18. Measurement of Moisture Storage Parameters of Building Materials

    Directory of Open Access Journals (Sweden)

    M. Jiřičková

    2003-01-01

    Full Text Available The moisture storage parameters of three different building materials: calcium silicate, ceramic brick and autoclaved aerated concrete, are determined in the hygroscopic range and overhygroscopic range. Measured sorption isotherms and moisture retention curves are then combined into moisture storage functions using the Kelvin equation. A comparison of measured results with global characteristics of the pore space obtained by mercury intrusion porosimetry shows a reasonable agreement; the median pore radii by volume are well within the interval given by the beginning and the end of the characteristic steep parts of the moisture retention curves.

  19. 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.

  20. Radiological characterisation of waste in interim storage building of COVRA

    International Nuclear Information System (INIS)

    At COVRA spatial dose rate distribution measurements were performed in December 2004 and December 2006 in the interim L/ILW storage building (LOG). This storage facility consists out of four large storage halls (height x width x depth 7 m x 40 m x 70 m) each with a volume of about 20000 m3. The scope of this study is to investigate the benefits of the waste storage strategy and procedures for minimization of the dose to the workers and the public. The main aim of the measurements in 2004 was: to validate the applied L/ILW storage strategy - to examine, if spatial collected data can be used to detect unforeseen differences in radiation level. The results of these measurements of spatial dose showed a number of unforeseen hotspots at different locations, so that it could be concluded that the applied storage strategy and procedures has to be improved. Further the dose rate at the height of 6 m, mainly responsible for the sky-shine dose rate, being an important part of the dose rate to the public at the site boundary, has to be reduced by more shielding. In December 2006 a second serial of spatial radiological and non-radiological data have been collected. The applied nondestructive INDSS-R (Indoor Survey System-Radiation ) method has been improved, so that the following 3-dimensional data could be collected between 0.5 m and 5.5 m: - dose rate (by pressurized ionisation chamber). nuclide depended gamma photon flux (3 x 3 NaI). - temperature and relative humidity. These last two non-radiological parameters were measured to verify the storage conditions of the waste. The main aim of these 3 dimensional collection was to verify the second stated aim of 2004. (authors)

  1. Integrated Building Energy Systems Design Considering Storage Technologies

    International Nuclear Information System (INIS)

    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

  2. Integrated Building Energy Systems Design Considering Storage Technologies

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-07

    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 projectperformed 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.

  3. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

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

  4. Staging and storage facility feasibility study. Final report

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  6. Review on thermal performance of phase change energy storage building envelope

    Institute of Scientific and Technical Information of China (English)

    WANG Xin; ZHANG YinPing; XlAO Wei; ZENG RuoLang; ZHANG QunLi; DI HongFa

    2009-01-01

    Improving the thermal performance of building envelope is an important way to save building energy consumption. The phase change energy storage building envelope is helpful to effective use of renewable energy, reducing building operational energy consumption, increasing building thermal comfort, and reducing environment pollution and greenhouse gas emission. This paper presents the concept of ideal energy-saving building envelope, which is used to guide the building envelope material selection and thermal performance design. This paper reviews some available researches on phase change building material and phase change energy storage building envelope. At last, this paper presents some current problems needed further research.

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

    Energy Technology Data Exchange (ETDEWEB)

    Schwinkendorf, K.N., Fluor Daniel Hanford

    1997-03-24

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially- built baskets containing either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the MCO/Cask, operations at the Cold Vacuum Drying Facility (CVDF), and storage in the Canister Storage Building (CSB). Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} @ 0.95 criticality safety limit.

  8. Safety evaluation of the Mixed Waste Storage Building (Building 643-43E)

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J.M.

    1992-01-27

    A safety evaluation has been conducted for the Mixed Waste Storage Building (MWSB) at the Savannah River Site. The results of this evaluation are compared with those contained in the Burial Ground Safety Analysis Report (SAR). The MWSB will function as an interim storage facility for Resource Conservation and Recovery Act (RCRA) regulated mixed waste. It will meet all applicable standards set forth by the Environmental Protection Agency (EPA), the South Carolina Department of Health and Environment Control (SCDHEC), and Department of Energy (DOE) Orders.

  9. Safety evaluation of the Mixed Waste Storage Building (Building 643-43E)

    International Nuclear Information System (INIS)

    A safety evaluation has been conducted for the Mixed Waste Storage Building (MWSB) at the Savannah River Site. The results of this evaluation are compared with those contained in the Burial Ground Safety Analysis Report (SAR). The MWSB will function as an interim storage facility for Resource Conservation and Recovery Act (RCRA) regulated mixed waste. It will meet all applicable standards set forth by the Environmental Protection Agency (EPA), the South Carolina Department of Health and Environment Control (SCDHEC), and Department of Energy (DOE) Orders

  10. Multi-Canister Overpack (MCO) Design Report

    International Nuclear Information System (INIS)

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

  11. Multi-canister overpack closure operations location study

    International Nuclear Information System (INIS)

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

  12. 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.

  13. RCRA closure of the Building 3001 Storage Canal

    International Nuclear Information System (INIS)

    The 3001 Storage Canal is located under portions of Buildings 3001 and 3019 at Oak Ridge National Laboratory (ORNL) and has a capacity of approximately 62,000 gallons of water. The term canal has historically been used to identify this structure, however, the canal is an in-ground reinforced concrete structure satisfying the regulatory definition of a tank. From 1943 through 1963, the canal in Building 3001 was designed to be an integral part of the system for handling irradiated fuel from the Oak Ridge Graphite Reactor. Because one of the main initial purposes of the reactor was to produce plutonium for the chemical processing pilot plant in Building 3019, the canal was designed to be the connecting link between the reactor and the pilot plant. During the war years, natural uranium slugs were irradiated in the reactor and then pushed out of the graphite matrix into the system of diversion plates and chutes which directed the fuel into the deep pit of the canal. After shutdown of the reactor, the canal was no longer needed for its designed purpose. Since 1964, the canal has only been used to store radioisotopes and irradiated samples under a water pool for radiation protection. This report describes closure alternatives

  14. Thermal Energy Storage in Building Integrated Thermal Systems: A Review. Part 2. Integration as Passive System

    OpenAIRE

    Niall, Dervilla; McCormack, Sarah; Griffiths, Philip; Cabeza, Luisa; Navarro, Lidia; Castell, Albert; de Grazia, Alvaro; Brown, Maria

    2015-01-01

    Energy consumption trends in residential and commercial buildings show a significant increase in recent decades. One of the key points for reducing energy consumption in buildings is to decrease the energy demand. Buildings envelopes are not just a structure they also provide protection from outdoor weather conditions always taking into account the local climate. Thermal energy storage has been used and applied to the building structure by taking advantage of sensible heat storage of material...

  15. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

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

  16. 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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-06-02

    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.

  19. 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.

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

    International Nuclear Information System (INIS)

    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

  1. 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.

  2. Improved Air-Treatment Canister

    Science.gov (United States)

    Boehm, A. M.

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.

  5. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

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

  6. 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

  7. Status report, canister fabrication

    International Nuclear Information System (INIS)

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

  8. 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

  9. K West Basin canister survey

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.

  14. Pressurization of whole element canister during staging

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    FARWICK, C.C.

    1999-12-09

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

  17. Dynamic simulation of residential buildings with seasonal sorption storage of solar energy - parametric analysis

    OpenAIRE

    Hennaut, Samuel; Thomas, Sébastien; Davin, Elisabeth; Andre, Philippe

    2011-01-01

    This work focuses on the evaluation of the performances of a solar combisystem coupled to seasonal thermochemical storage using SrBr2/H20 as adsorbent/adsorbate couple. The objective is to determine the characteristics required for solar system and storage reactor to reach a 100 % solar fraction for a building with a low heating load. The complete system, including the storage reactor, is simulated, using the dynamic simulation software TRNSYS. The influence of some components and p...

  18. Storage facility for radioactive wastes

    International Nuclear Information System (INIS)

    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.)

  19. The Indonesia Carbon Capture Storage Capacity Building Program

    OpenAIRE

    World Bank

    2015-01-01

    In order to meet the growing Indonesian demand for electricity, while also constraining carbon dioxide (CO2) emissions, future coal power plants may have to include CO2 capture equipment with storage of that CO2. This study set out to define and evaluate the conditions under which fossil fuel power plants can be deemed as carbon capture and storage (CCS) ready (CCS-R). It considers the tec...

  20. Hashemite Kingdom of Jordan : Carbon Capture and Storage Capacity Building Technical Assistance

    OpenAIRE

    World Bank

    2012-01-01

    This study was funded by the Carbon Capture and Storage (CCS) capacity building trust fund, and administered by the World Bank. The main objectives of the study were: to build or enhance Jordan s institutional capacity to make informed policy decisions on CCS technology and applications; to assess the potential application of CCS technology in Jordan; and to identify barriers-legal, regula...

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

    International Nuclear Information System (INIS)

    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.

  2. Moisture insensitive charcoal canisters

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Meling, T.A.

    1995-03-28

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

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

    International Nuclear Information System (INIS)

    The following separate reports and correspondence pertains to the independent review of the seismic analysis. The original analysis was performed by GEC-Alsthom Engineering Systems Limited (GEC-ESL) under subcontract to Foster-Wheeler Environmental Corporation (FWEC) who was the prime integration contractor to the Spent Nuclear Fuel Project for the Multi-Canister Overpack (MCO) Handling Machine (MHM). The original analysis was performed to the Design Basis Earthquake (DBE) response spectra using 5% damping as required in specification, HNF-S-0468 for the 90% Design Report in June 1997. The independent review was performed by Fluor-Daniel (Irvine) under a separate task from their scope as Architect-Engineer of the Canister Storage Building (CSB) in 1997. The comments were issued in April 1998. Later in 1997, the response spectra of the Canister Storage Building (CSB) was revised according to a new soil-structure interaction analysis and accordingly revised the response spectra for the MHM and utilized 7% damping in accordance with American Society of Mechanical Engineers (ASME) NOG-1, ''Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).'' The analysis was re-performed to check critical areas but because manufacturing was underway, designs were not altered unless necessary. FWEC responded to SNF Project correspondence on the review comments in two separate letters enclosed. The dispositions were reviewed and accepted. Attached are supplier source surveillance reports on the procedures and process by the engineering group performing the analysis and structural design. All calculation and analysis results are contained in the MHM Final Design Report which is part of the Vendor Information File 50100. Subsequent to the MHM supplier engineering analysis, there was a separate analyses for nuclear safety accident concerns that used the electronic input data files provided by FWEC/GEC-ESL and are contained in document SNF-6248

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gregor P. Henze; Moncef Krarti

    2003-12-17

    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 investigates the merits of harnessing both storage media concurrently in the context of predictive optimal control. This topical report describes the demonstration of the model-based predictive optimal control for active and passive building thermal storage inventory in a test facility in real-time using time-of-use differentiated electricity prices without demand charges. The laboratory testing findings presented in this topical report cover the second of three project phases. The novel supervisory controller successfully executed a three-step procedure consisting of (1) short-term weather prediction, (2) optimization of control strategy over the next planning horizon using a calibrated building model, and (3) post-processing of the optimal strategy to yield a control command for the current time step that can be executed in the test facility. The primary and secondary building mechanical systems were effectively orchestrated by the model-based predictive optimal controller in real-time while observing comfort and operational constraints. The findings reveal that when the optimal controller is given imperfect weather fore-casts and when the building model used for planning control strategies does not match the actual building perfectly, measured utility costs savings relative to conventional building operation can be substantial. This requires that the facility under control lends itself to passive storage utilization and the building

  7. 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.

  8. Drying behavior of K-East canister sludge

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  10. 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

  11. Characterization of polymers and Microencapsulated Phase Change Materials used for Thermal Energy Storage in buildings

    OpenAIRE

    Giró Paloma, Jessica

    2015-01-01

    [eng] The use of renewable heat decreases the consumption of fossil resources, although its usage is intermittent and usually does not match the demand. A proper thermal energy storage system design can eliminate this problem by reducing the consumption of non-renewable resources and improving energy efficiency where used. In buildings, thermal energy storage using phase change materials (PCM) is a useful tool to achieve reduction in energy consumption. These can be incorporated into passive ...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  14. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

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

  15. 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.

  16. 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.

  17. Monitoring water storage variations in the vadose zone with gravimeters - quantifying the influence of observatory buildings

    Science.gov (United States)

    Reich, Marvin; Güntner, Andreas; Mikolaj, Michal; Blume, Theresa

    2016-04-01

    Time-lapse ground-based measurements of gravity have been shown to be sensitive to water storage variations in the surroundings of the gravimeter. They thus have the potential to serve as an integrative observation of storage changes in the vadose zone. However, in almost all cases of continuous gravity measurements, the gravimeter is located within a building which seals the soil beneath it from natural hydrological processes like infiltration and evapotranspiration. As water storage changes in close vicinity of the gravimeter have the strongest influence on the measured signal, it is important to understand the hydrology in the unsaturated soil zone just beneath the impervious building. For this reason, TDR soil moisture sensors were installed in several vertical profiles up to a depth of 2 m underneath the planned new gravimeter building at the Geodetic Observatory Wettzell (southeast Germany). In this study, we assess the influence of the observatory building on infiltration and subsurface flow patterns and thus the damping effect on gravimeter data in a two-way approach. Firstly, soil moisture time series of sensors outside of the building area are correlated with corresponding sensors of the same depth beneath the building. The resulting correlation coefficients, time lags and signal to noise relationships are used to find out how and where infiltrating water moves laterally beneath the building and towards its centre. Secondly, a physically based hydrological model (HYDRUS) with high discretization in space and time is set up for the 20 by 20 m area around and beneath the gravimeter building. The simulated spatial distribution of soil moisture in combination with the observed point data help to identify where and to what extent water storage changes and thus mass transport occurs beneath the building and how much this differs to the dynamics of the surroundings. This allows to define the umbrella space, i.e., the volume of the vadose zone where no mass

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. 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

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

    International Nuclear Information System (INIS)

    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 ft2 of space in the 2101-M Building in the 200 East Area. Here, modified storage space is estimated to cost about $106 per ft2 while new construction will cost about $50 per ft2. 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

  2. Superhalogens as Building Blocks of Complex Hydrides for Hydrogen Storage

    CERN Document Server

    Srivastava, Ambrish Kumar

    2016-01-01

    Superhalogens are species whose electron affinity (EA) or vertical detachment energy (VDE) exceed to those of halogen. These species typically consist of a central electropositive atom with electronegative ligands. The EA or VDE of species can be further increased by using superhalogen as ligands, which are termed as hyperhalogen. Having established BH4- as a superhalogen, we have studied BH4-x(BH4)x- (x = 1 to 4) hyperhalogen anions and their Li-complexes, LiBH4-x(BH4)x using density functional theory. The VDE of these anions is larger than that of BH4-, which increases with the increase in the number of peripheral BH4 moieties (x). The hydrogen storage capacity of LiBH4-x(BH4)x complexes is higher but binding energy is smaller than that of LiBH4, a typical complex hydride. The linear correlation between dehydrogenation energy of LiBH4-x(BH4)x complexes and VDE of BH4-x(BH4)x- anions is established. These complexes are found to be thermodynamically stable against dissociation into LiBH4 and borane. This stud...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. 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.

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

    OpenAIRE

    Alexandre Hugo; Radu Zmeureanu

    2012-01-01

    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 softwar...

  8. 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.

  9. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  12. Energy Efficiency through Thermal Energy Storage - Evaluation of the Possibilities for the Swedish Building Stock, Phase 1

    OpenAIRE

    Heier, Johan; Bales, Chris; Martin, Viktoria

    2010-01-01

    As a first step in assessing the potential of thermal energy storage in Swedish buildings, the current situation of the Swedish building stock and different storage methods are discussed in this paper. Overall, many buildings are from the 1960’s or earlier having a relatively high energy demand, creating opportunities for large energy savings. The major means of heating are electricity for detached houses and district heating for multi dwelling houses and premises. Cooling needs are relativel...

  13. Long term Integrity of Spent Fuel and Construction Materials and Behaviour of Components for Dry Storage Facilities

    International Nuclear Information System (INIS)

    In Japan, two dry storage facilities at TEPCO and JAPCO sites have been in operation since 1995 and 2002 respectively. The TEPCO dry storage facility was damaged by a Tsunami attack on 11th March 2011. The casks stored in the facility have since been moved to an onsite temporary cask custody area; after confirmation of the integrity of casks. On the other hand, the Tsunami did not attack the dry storage facility at the JAPCO site. The integrity of the storage building and the casks were maintained. In addition, an off-site centralized dry storage facility has been constructed at Mutsu City. Operation of the storage facility is pending a safety re-examination against new safety regulations. Its final storage capacity will be 5000 t(U) and the storage period is up to 50 years. To support storage operations, it is therefore necessary to obtain and evaluate data on the integrity of spent fuels and cask construction materials during long term dry storage. Objectives: Construction materials for dry storage facilities: • To evaluate long-term reliability of welded stainless steel canisters under stress corrosion cracking (SCC) environment, including the critical salt density deposited on the canister to initiate SCC, monitoring, prevention, and the mitigation method of SCC; • To detect and analyze the cover gas leak from canisters; • To evaluate integrity of sealability of metal gasket under long term storage; • To evaluate influence of the vibration on sealing performance of the ageing gasket

  14. Storage technologies of spent fuel. For the stable operation of nuclear fuel cycle

    International Nuclear Information System (INIS)

    Spent fuel requires an interim storage under dry conditions until reprocessing. This paper reports the dry storage technologies. The main dry storage technologies include three methods: metal cask storage, concrete cask storage, and vault storage. The metal cask storage and vault storage have been adopted in Japan. Metal cask is a cylindrical container made of steel, and has four safety functions such as prevention of criticality, confinement, heat removal, and shielding. The dimensions of the largest one are about 2.5 m in diameter and 5 m in height, weighing about 120 tons. In the concrete cask storage, spent fuel is placed in a thin-walled cylindrical container called canister, and the canister is placed in a concrete reservoir in a vertical posture and stored. The concrete cask has a weight of about 180 tons. The atmosphere in the container is helium gas, the same as the case of the metal cask. In the vault storage, the canisters filled with spent fuel are stored side by side in a large space of concrete building, and cooled by air flow. As common challenge, there is the establishment of a thermal analysis technique for evaluating the aging deterioration and soundness of the structural materials. (A.O.)

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

    International Nuclear Information System (INIS)

    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 RME3 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)

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

    International Nuclear Information System (INIS)

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

  17. The integration of water loop heat pump and building structural thermal storage systems

    Energy Technology Data Exchange (ETDEWEB)

    Marseille, T.J.; Schliesing, J.S.

    1991-10-01

    Many commercial buildings need heat in one part and, at the same time, cooling in another part. Even more common is the need for heating during one part of the day and cooling during another in the same spaces. If that energy could be shifted or stored for later use, significant energy might be saved. If a building's heating and cooling subsystems could be integrated with the building's structural mass and used to collect, store, and deliver energy, the energy might be save cost-effectively. To explore this opportunity, researchers at the Pacific Northwest Laboratory (PNL) examined the thermal interactions between the heating, ventilating, and air-conditioning (HVAC) system and the structure of a commercial building. Computer models were developed to simulate the interactions in an existing building located in Seattle, Washington, to determine how these building subsystems could be integrated to improve energy efficiency. The HVAC subsystems in the existing building were modeled. These subsystems consist of decentralized water-source heat pumps (WSHP) in a closed water loop, connected to cooling towers for heat rejection during cooling mode and boilers to augment heating. An initial base case'' computer model of the Seattle building, as-built, was developed. Metered data available for the building were used to calibrate this model to ensure that the analysis would provide information that closely reflected the operation of a real building. The HVAC system and building structure were integrated in the model using the concrete floor slabs as thermal storage media. The slabs may be actively charged during off-peak periods with the chilled water in the loop and then either actively or passively discharged into the conditioned space during peak periods. 21 refs., 37 figs., 17 tabs.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Optimisation of combined heat and power production for buildings using heat storage

    International Nuclear Information System (INIS)

    Highlights: • Half-hourly heat demand data shows the high variability of building heat demand. • Sharp spikes in heat demand were observed when some heating systems are activated. • 25% of the annual heat demand was found to be independent of outdoor temperatures. • Seasonal differences of heat store operation affect its environmental and economic advantages. - Abstract: Reducing carbon emissions from buildings is vital to achieve goals for avoiding dangerous climate change, and supplying them with low-carbon heat is essential. In the UK, the development of heat networks for supplying low-carbon heat is being encouraged for urban areas where there is high heat demand density. This paper investigates heat demand variability, the role of heat networks and combined heat and power (CHP) in satisfying this demand, and finally the advantages of using heat storage in the system. Building heat demands from 50 buildings were analysed at a half-hour resolution with modelling to determine CHP operation patterns with and without heat storage. Daily total heat demand was found to vary from 25% of the full-year average in summer months up to 235% of the average on the coldest days in winter. The heat demand was shown to correlate to outdoor temperatures measured with the degree-day parameter, except for approximately 100 days during the warmest part of the year falling outside the heating season. Sharp spikes in heat demand were seen at the half-hourly time scale coinciding with the switching on of heating systems in some buildings with consequences for building energy supply options. It was shown that for an annual heat demand of 40,000 MW h, the use of thermal storage can significantly increase the running time of a CHP energy centre with 4 MW capacity designed to supply this demand. The cost savings resulting from increased on-site heat and electricity production resulted in a payback period for heat storage investment of under four years with further benefits if it

  1. 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 and...... reliable means for seasonal storage of thermal energy. This is particularly true at locations where seasonal variations of solar radiation are significant and/or in climates where seasonally varying space heating and cooling loads dominate energy consumption. This article conducts a literature review of......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...

  2. 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...... (and the the return temperature) would only be a few degrees above room temperature due to the very low heating demand and the large heat transfer surface area. One of the objectives in a newly started IEA Task 32 project is to investigate and develop improved thermal storages for combined solar...... a stable super cooling, i.e. the material is able to cool down below its freezing point (Tfusion) and still be liquid, the possibility exist for a storage with a very low heat loss. When energy is needed from the storage the solidification is activated and the temperature rises almost instantly to...

  3. Cold Thermal Storage and Peak Load Reduction for Office Buildings in Saudi Arabia

    Institute of Scientific and Technical Information of China (English)

    Nabil Y.Abdel-Shafi; Ramzy R.Obaid; Ibrahim M.Jomoah

    2014-01-01

    This paper involves the investigations of the chilled water and ice cold thermal storage technologies along with the associated operating strategies for the air conditioning (AC) systems of the typical office buildings in Saudi Arabia, so as to reduce the electricity energy consumption during the peak load periods. In Saudi Arabia, the extensive use of AC for indoor cooling in offices composes a large proportion of the annual peak electricity demand. The very high temperatures over long summer periods, extending from May to October, and the low cost of energy are the key factors in the wide and extensive use of air conditioners in the kingdom. This intense cooling load adds up to the requirement increase in the capacity of power plants, which makes them under utilized during the off-peak periods. Thermal energy storage techniques are one of the effective demand-side energy management methods. Systems with cold storage shifts all or part of the electricity requirement from peak hours to off-peak hours to reduce demand charges and/or take advantage of off-peak rates. The investigations reveal that the cold thermal energy storage techniques are effective from both technical and economic perspectives in the reduction of energy consumption in the buildings during peak periods.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    DEFF Research Database (Denmark)

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

    2007-01-01

    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...... potential. However, because heat gains and night ventilation periods do not coincide in time, a sufficient amount of thermal mass is needed in the building to store the heat. Assuming a 24 h-period harmonic oscillation of the indoor air temperature within a range of thermal comfort, the analytical solution...... 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...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

    Gas and liquid samples have been obtained from a selection of the approximately 3,820 spent fuel storage canisters in the K West Basin. The samples were taken to characterize the contents of the gas and water in the canisters. The data will provide source term information for two subprojects of the Spent Nuclear Fuel Project (SNFP) (Fulton 1994): the K Basins Integrated Water Treatment System subproject (Ball 1996) and the K Basins Fuel Retrieval System subproject (Waymire 1996). The barrels of ten canisters were sampled in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results for the gas and liquid samples of the first campaign have been reported (Trimble 1995a; Trimble 1995b; Trimble 1996a; Trimble 1996b). An analysis of cesium-137 (137CS ) data from the second campaign samples was reported (Trimble and Welsh 1997), and the gas sample results are documented in Trimble 1997. This report documents the results of all analytes of liquid samples from the second campaign.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  12. 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)

    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

  13. 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.

  14. 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

  15. Canister compatibility with Carlsbad salt

    International Nuclear Information System (INIS)

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

  16. Canister Transfer Facility Criticality Calculations

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Monroe-Rammsy

    2000-10-13

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  19. Performance evaluation of solar-assisted air-conditioning system with chilled water storage (CIESOL building)

    International Nuclear Information System (INIS)

    Highlights: ► We present a new solar-assisted air-conditioning system’s operation sequence. ► This mode considers the chilled water tanks action with variable-speed pump. ► It permits to save about 20% and 30% of energy and water consumption, respectively. ► It allows storing the excess cooling capacity of the absorption chiller. ► It prevents the sudden start/stop (on/off cycles) of the absorption chiller. - Abstract: This study presents the performance of solar-assisted air-conditioning system with two chilled water storage tanks installed in the Solar Energy Research Center building. The system consists mainly of solar collectors’ array, a hot-water driven absorption chiller, a cooling tower, two hot storage tanks, an auxiliary heater as well as two chilled storage tanks. The chilled water storage tank circuit was further investigated in order to find the optimum solar system’s operation sequence while providing the best energy performance. Firstly, we carried out a study about the dynamics of building’s cooling load and the necessity of the integration of chilled water storage tanks to solar system. Subsequently, the new system’s operation mode was proposed to reduce the energy consumption. The results demonstrate that we can save about 20% of the total energy consumption and about 30% of water consumption applying the new operation sequence, which takes into account the chilled water tanks action. Moreover, it was demonstrated that the integration of chilled water storage tanks allows to reduce the sudden absorption chiller on/off cycles, thereby improving the efficiency of the solar-assisted system.

  20. 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 ta...... 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.......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...

  1. 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

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

    International Nuclear Information System (INIS)

    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%

  3. Preparation and characterization of phase change material for thermal energy storage in buildings

    Science.gov (United States)

    Lo, Tommy Y.

    2016-04-01

    The paper presents the developing of novel form-stable composite phase change material (PCM) by incorporation of paraffin into lightweight aggregate through vacuum impregnation. The macro-encapsulated Paraffin-lightweight aggregate is a chemical compatible, thermal stable and thermal reliable PCM material for thermal energy storage applications in buildings. The 28 days compressive strength of NWAC using PCM-LWA is 33 - 53 MPa, which has an opportunity for structural purpose. Scanning electronic microscopic images indicated the paraffin can be held inside the porous structure of the aggregate. Thermal performance test showed that the cement paste panel with composite PCM can reduce the indoor temperature.

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

    International Nuclear Information System (INIS)

    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

  5. Criticality safety studies of Building 3019 Cell 4 and in-line storage wells

    International Nuclear Information System (INIS)

    New fissile material load limits for storage facilities located in Building 3019 are derived in a manner consistent with currently applicable Martin Marietta Energy Systems requirements. The limits for 233U loading are 2.00, 1.80, 1.45, and 0.19 kg/ft for hydrogen-to-233U atoms ratios of 3, 5, 10, and unrestricted, respectively. Limits were also found for 235U and 239Pu systems. The KENO-Va Monte Carlo Program and Hansen-Roach cross sections were used to derive these limits

  6. Designing And Testing An Air-PCM Heat Exchanger For Building Ventilation Application Coupled To Energy Storage

    OpenAIRE

    Dechesne, Bertrand; Gendebien, Samuel; Martens, Jonathan; Lemort, Vincent

    2014-01-01

    Due to the increase of energy costs, buildings energy consumption has tended to decrease in the past decades. This gives an opportunity for developing innovative renewable technologies that are more adapted to recent buildings with low energy demand. In this context, one main challenge is to manage non-simultaneous availability of heat source or sink and the energy demand of buildings. Hence, different technologies dedicated to energy storage have been developed recently; one of them is the u...

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

    International Nuclear Information System (INIS)

    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

  8. Retrievability of spent nuclear fuel canisters

    International Nuclear Information System (INIS)

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

  9. 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.

  10. 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.

  11. 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.

  12. Safety evaluation for the inner canister closure station

    International Nuclear Information System (INIS)

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

  13. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

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

  17. 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.

  18. 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.

  19. Device for storage of radioactive material in a building with heat pipes set in the building wall

    International Nuclear Information System (INIS)

    When storing radio-active material in a building, safe and sufficient heat removal must always be guaranteed. On the other hand, the building should be safely closed to the environment. The invention makes it possible to ensure, for such a building with heat pipes set in the building wall, that it is possible to use at least part of the heat generated in the building without limiting the removal of heat. Cooling sleeves are fitted to the heat pipes near the building wall for this purpose, where a cooling circuit with a circulating coolant is connected to the cooling sleeves. (orig.)

  20. 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.

  1. Analysis of sludge from Hanford K East Basin canisters

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-05-15

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 micro

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

    International Nuclear Information System (INIS)

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

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  9. Canister transfer into repository in shaft alternative

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R; Enos, David

    2014-12-01

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

  11. Conceptual designs of radioactive canister transporters

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-02-01

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

  12. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Statistical analysis of DWPF reference canister dimensions

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  17. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

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

  18. Site status monitoring report for underground storage tank 2331-U at Building 9201-1

    International Nuclear Information System (INIS)

    The purpose of this document is to present potentiometric, groundwater quality and vapor monitoring data required for site status monitoring of underground storage tank (UST) 2331-U at the Building 9201-1 Site. Site status monitoring has been conducted at the site as part of a Monitoring Only program approved by the Tennessee Department of Environment and Conservation (TDEC) based on review and approval of Site Ranking (Site Ranking Form approved May 23, 1994). This document presents the results of the first semiannual site status monitoring that was performed in December 1994. Site status monitoring and preparation of this report have been conducted in accordance with the requirements of TDEC Rule 1200-1-15 and the TDEC UST Reference Handbook, Second Edition (TDEC 1994) Technical Guidance Document (TGD) 007. This document is organized into three sections. Section 1 presents introductory information relative to the site including the regulatory initiative and a site description. Section 2 includes the results of measurement and sampling of monitoring wells GW-193, GW-657, GW-707, GW-708, GW-808, GW-809, and GW-810. Section 3 presents data from vapor monitoring conducted in subsurface utilities present at the site

  19. 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.

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

    International Nuclear Information System (INIS)

    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 103 kV/m the particle collection rate could be improved dramatically

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  3. Technical Support Document: The Development of the Advanced Energy Design Guide for Small Warehouse and Self-Storage Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Bing; Jarnagin, Ronald E.; Jiang, Wei; Gowri, Krishnan

    2007-12-01

    This Technical Support Document (TSD) describes the process and methodology for development of the Advanced Energy Design Guide for Small Warehouse and Self-storage Buildings (AEDG-WH or the Guide), a design guidance document intended to provide recommendations for achieving 30% energy savings in small warehouses over levels contained in ANSI/ASHRAE/IESNA Standard 90.1-1999, Energy Standard for Buildings Except Low-Rise Residential Buildings. The AEDG-WH is the fourth in a series of guides being developed by a partnership of organizations, including the American Society of Heating, Refrigerating and Air-Conditioning Engineers, Inc. (ASHRAE), the American Institute of Architects (AIA), the Illuminating Engineering Society of North America (IESNA), the United States Green Buildings Council (USGBC), and the U.S. Department of Energy (DOE).

  4. 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.

    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).

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

    International Nuclear Information System (INIS)

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

  6. 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

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

    International Nuclear Information System (INIS)

    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)

  8. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

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

  9. 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.)

  10. 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.

  11. Remote controlled mover for disposal canister transfer

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    CALMUS, R.B.

    2000-12-27

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

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

    International Nuclear Information System (INIS)

    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

  14. 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...... the performance of the new deck with PCM concrete is the thermal properties of such a new material, as the PCM concrete is yet to be well defined. The results presented in the paper include models in which the PCM concrete material properties, such as thermal conductivity, and specific heat capacity were first...... with and without microencapsulated PCM are presented. The new concrete deck with microencapsulated PCM is the standard deck on which an additional layer of the PCM concrete was added and, at the same time, the latent heat storage was introduced to the construction. The challenge of numerically simulating...

  15. 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.

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

    International Nuclear Information System (INIS)

    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

  17. 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 and...... xanthane rubber. The storage can cool down to surrounding temperature preserving the latent heat in form of the heat of fusion energy. The basis for the calculations is a super low energy house with a space heating demand of 2010 kWh/year and a domestic hot water demand of 2530 kWh/year. For storage...... volumes in the range of 500 – 3000 litres the heat loss free state is seldom reached and the effect of super cooling is limited. For larger volumes the heat loss free state may be reached. The benefit of using a PCM storage compared to a traditional water storage is limited with respect to energy savings...

  18. 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

    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 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...

  19. Damaged Fuel Storage and Recovery - A Case Study

    International Nuclear Information System (INIS)

    Since the 1960s, spent nuclear fuel section test pieces and other severely damaged nuclear fuel sections from a variety of sources was placed into cans within large isolation canisters, termed Oversize (OS) canisters, for interim wet storage at the Savannah River Site (SRS) Receiving Basin for Offsite Fuel (RBOF). The water activity in the OS canisters had risen with time to extremely high levels (5.8E6 Bq/ml in canister A5) and the canisters also contained corrosion product sludge debris. The fuel cans from the RBOF OS canisters were destined to be removed and transferred to the SRS L-Basin to consolidate fuel storage at the site and complete the fuel deinventory of RBOF. A special underwater filter/deionizer apparatus was designed, constructed, and deployed to efficiently capture the cesium from the OS canister water prior to opening the OS canister and to prevent it from dispersal into the general RBOF basin water and mitigate high radiation levels above the water. New OS canisters for L-basin were designed to hold the cans of the damaged fuel from RBOF. The new OS canisters included updated features for ease of handling and storage in racks in L-basin. The new canisters included a modified vent path from the canister to the general basin water to allow hydrogen to safely escape from the corroding fuel while containing high-activity water with the canister. This case study will relate experiences with containment of damaged fuels in pool storage and mitigation of effects of any potential cesium release into the water. (author)

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

    International Nuclear Information System (INIS)

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

  1. On scale and magnitude of pressure build-up induced by large-scale geologic storage of CO2

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Q.; Birkholzer, J. T.

    2011-05-01

    The scale and magnitude of pressure perturbation and brine migration induced by geologic carbon sequestration is discussed assuming a full-scale deployment scenario in which enough CO{sub 2} is captured and stored to make relevant contributions to global climate change mitigation. In this scenario, the volumetric rates and cumulative volumes of CO{sub 2} injection would be comparable to or higher than those related to existing deep-subsurface injection and extraction activities, such as oil production. Large-scale pressure build-up in response to the injection may limit the dynamic storage capacity of suitable formations, because over-pressurization may fracture the caprock, may drive CO{sub 2}/brine leakage through localized pathways, and may cause induced seismicity. On the other hand, laterally extensive sedimentary basins may be less affected by such limitations because (i) local pressure effects are moderated by pressure propagation and brine displacement into regions far away from the CO{sub 2} storage domain; and (ii) diffuse and/or localized brine migration into overlying and underlying formations allows for pressure bleed-off in the vertical direction. A quick analytical estimate of the extent of pressure build-up induced by industrial-scale CO{sub 2} storage projects is presented. Also discussed are pressure perturbation and attenuation effects simulated for two representative sedimentary basins in the USA: the laterally extensive Illinois Basin and the partially compartmentalized southern San Joaquin Basin in California. These studies show that the limiting effect of pressure build-up on dynamic storage capacity is not as significant as suggested by Ehlig-Economides and Economides, who considered closed systems without any attenuation effects.

  2. Concept Evaluation of Pyroprocess Waste Storage System

    International Nuclear Information System (INIS)

    The wastes should be fabricated into stable waste forms considering their radioactivity and decay heat characteristics for the safety management of long-term storage and final disposal. The purpose of this study is to assess the concept of a waste storage canister and facility for securing the design data. The design concepts of the waste form storage canister and facility were developed for the storage of pyroprocess waste. The optimal waste form dimension was determined as Φ350 x 900 mm(H) with a maximum decay heat of 1.9 kW. In addition, the canister concept was proposed with an outer dimension ofΦ 370 x 1320 mm(H). The concept of an integrated storage facility was proposed for a vault storage system. It shows that the thermal integrity of the storage facility was maintained under normal condition. In addition, it was found that the cooling efficiency of the storage facility is superior

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

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, J.M.; Furbo, S. [Technical Univ. of Denmark, Dept. of Civil Engineering, Kgs.Lyngby (Denmark)

    2006-05-04

    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 and xanthane rubber. The storage can cool down to surrounding temperature preserving the latent heat in form of the heat of fusion energy. The basis for the calculations is a super low energy house with a space heating demand of 2010 kWh/year and a domestic hot water demand of 2530 kWh/year. For storage volumes in the range of 500 3000 litres the heat loss free state is seldom reached and the effect of super cooling is limited. For larger volumes the heat loss free state may be reached. The benefit of using a PCM storage compared to a traditional water storage is limited with respect to energy savings for storage sizes up to 1 m{sup 3}, but if the same amount of net utilised solar energy should be reached it would require a water storage that is 2 3 times larger. (au)

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

    International Nuclear Information System (INIS)

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

  5. 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)

  6. 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...... be self-made to keep the price down. The system is working, but heat exchange from plastic piping to sand is rather poor. The dimensioning of the volume is rather difficult based on common knowledge. Passive heating, hence reduction of heat demand, due to the storage and especially due to the oversized...

  7. Rehearsal: Sample Canister in Cleanroom (Animation)

    Science.gov (United States)

    2005-01-01

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

  8. Techniques for freeing deposited canisters. Final report

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    2001-01-01

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

  10. 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,…

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sreten Mastilovic

    2001-07-31

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

  13. Health risk assessment for the Building 3001 Storage Canal at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    This human health risk assessment has been prepared for the Environmental Restoration (ER) Program at the Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee. The objectives of this risk assessment are to evaluate the alternatives for interim closure of the Building 3001 Storage Canal and to identify the potential health risk from an existing leak in the canal. The Building 3001 Storage Canal connects Buildings 3001 and 3019. The volume of water in the canal is monitored and kept constant at about 62,000 gal. The primary contaminants of the canal water are the radionuclides 137Cs, 60Co, and 90Sr; a layer of sediment on the canal floor also contains radionuclides and metals. The prime medium of contaminant transport has been identified as groundwater. The primary route for occupational exposure at the canal is external exposure to gamma radiation from the canal water and the walls of the canal. Similarly, the primary exposure route at the 3042 sump is external exposure to gamma radiation from the groundwater and the walls of the sump. Based on the exposure rates in the radiation work permits (Appendix C) and assuming conservative occupational work periods, the annual radiation dose to workers is considerably less than the relevant dose limits. The potential risk to the public using the Clinch River was determined for three significant exposure pathways: ingestion of drinking water; ingestion of contaminated fish; and external exposure to contaminated sediments on the shoreline, the dominant exposure pathway

  14. Optimization of a thermal storage unit combined with a biomass bioler for heating buildings

    OpenAIRE

    Butala, Vincenc; Stritih, Uroš

    2015-01-01

    The performance of a boiler with a built-in thermal storage unit is presented.The thermal storage unit is an insulated water tank that absorbs surplus heat from the boiler. The stored heat in the thermal storage unit makes it possible to heat even when the boiler is not operating, thus increasing the heating efficiency. A system with three components is described. The model of the system and the mathematical model were made using the TRNSYS program package and a test reference year (TRY). The...

  15. 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.

  16. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    The report provides a summary of the design of the canister for final disposal of nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  17. Modelling of solar thermo-chemical system for energy storage in buildings

    OpenAIRE

    Skrylnyk, Alexandre; Courbon, Emilie; Frère, Marc; Hennaut, Samuel; Andre, Philippe; Sun, Philippe; Descy, Gilbert

    2012-01-01

    The goal of this paper is the demonstration of the methodological design principles within theoretical modelling of thermal heat storage apparatus and simulation of inter-seasonal heat storage system. The designing procedure starts from the modelling of thermal plant behaviour, based on the simplifications in the basic hypothesis. Afterwards, a more detailed modelling, involving dynamic aspects and additional features of plant components, is prese...

  18. Initial findings: The integration of water loop heat pump and building structural thermal storage systems

    Energy Technology Data Exchange (ETDEWEB)

    Marseille, T.J.; Johnson, B.K.; Wallin, R.P.; Chiu, S.A.; Crawley, D.B.

    1989-01-01

    This report is one in a series of reports describing research activities in support of the US Department of Energy (DOE) Commercial Building System Integration Research Program. The goal of the program is to develop the scientific and technical basis for improving integrated decision-making during design and construction. Improved decision-making could significantly reduce buildings' energy use by the year 2010. The objectives of the Commercial Building System Integration Research Program are: to identify and quantify the most significant energy-related interactions among building subsystems; to develop the scientific and technical basis for improving energy related interactions in building subsystems; and to provide guidance to designers, owners, and builders for improving the integration of building subsystems for energy efficiency. The lead laboratory for this program is the Pacific Northwest Laboratory. A wide variety of expertise and resources from industry, academia, other government entities, and other DOE laboratories are used in planning, reviewing and conducting research activities. Cooperative and complementary research, development, and technology transfer activities with other interested organizations are actively pursued. In this report, the interactions of a water loop heat pump system and building structural mass and their effect on whole-building energy performance is analyzed. 10 refs., 54 figs., 1 tab.

  19. 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...... energy and payback period. It is shown that the battery size leading to the minimum payback period within the input range, is comprised between 2.6 kWh and 6.2 kWh. The lowest payback periods, (~5.6 years), are reached with a well-insulated building envelope, a high lightning and appliance consumption, a...... 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...

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

    International Nuclear Information System (INIS)

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

  1. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 m3 (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 m2. The heat collection unit consisted of 5 m2 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

  4. Dry storage systems with free convection air cooling

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  6. Fatman:Building Reliable Archival Storage Based on Low-Cost Volunteer Resources

    Institute of Scientific and Technical Information of China (English)

    覃安; 胡殿明; 刘俊; 杨文君; 谭待

    2015-01-01

    We present Fatman, an enterprise-scale archival storage based on volunteer contribution resources from under-utilized web servers, usually deployed on thousands of nodes with spare storage capacity. Fatman is specifically designed for enhancing the utilization of existing storage resources and cutting down the hardware purchase cost. Two ma jor con-cerned issues of the system design are maximizing the resource utilization of volunteer nodes without violating service level objectives (SLOs) and minimizing the cost without reducing the availability of archival system. Fatman has been widely deployed on tens of thousands of server nodes across several datacenters, providing more than 100 PB storage capacity and serving dozens of internal mass-data applications. The system realizes an efficient storage quota consolidation by strong isolation and budget limitation, to maximally support resource contribution without any degradation on host-level SLOs. It novelly improves data reliability by applying disk failure prediction to minish failure recovery cost, named fault-aware data management, dramatically reduces the mean time to repair (MTTR) by 76.3% and decreases file crash ratio by 35%on real-life product workload.

  7. Modeling And Predictive Control Of High Performance Buildings With Distributed Energy Generation And Thermal Storage

    OpenAIRE

    Li, Siwei; Karava, Panagiota

    2014-01-01

    Building-integrated photovoltaic-thermal (BIPV/T) systems replace conventional building cladding with solar technology that generates electricity and heat. For example, unglazed transpired solar collectors, known as UTCs, can be integrated with open-loop photovoltaic thermal (PV/T) systems to preheat ventilation air and/or to feed hot air into an air source heat pump, thus satisfying a significant part of the building’s heating and/or hot water requirements while also generating electricity. ...

  8. The integration of water loop heat pump and building structural thermal storage systems

    Energy Technology Data Exchange (ETDEWEB)

    Marseille, T.J.; Schliesing, J.S.

    1990-09-01

    Commercial buildings often have extensive periods where one space needs cooling and another heating. Even more common is the need for heating during one part of the day and cooling during another in the same spaces. If a building's heating and cooling system could be integrated with the building's structural mass such that the mass can be used to collect, store, and deliver energy, significant energy might be saved. Computer models were developed to simulate this interaction for an existing office building in Seattle, Washington that has a decentralized water-source heat pump system. Metered data available for the building was used to calibrate a base'' building model (i.e., nonintegrated) prior to simulation of the integrated system. In the simulated integration strategy a secondary water loop was manifolded to the main HVAC hydronic loop. tubing in this loop was embedded in the building's concrete floor slabs. Water was routed to this loop by a controller to charge or discharge thermal energy to and from the slabs. The slabs were also in thermal communication with the conditioned spaces. Parametric studies of the building model, using weather data for five other cities in addition to Seattle, predicted that energy can be saved on cooling dominated days. On hot, dry days and during the night the cooling tower can beneficially be used as a free cooling'' source for thermally charging'' the floor slabs using cooled water. Through the development of an adaptive/predictive control strategy, annual HVAC energy savings as large as 30% appear to be possible in certain climates. 8 refs., 13 figs.

  9. 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

  10. CANISTER HANDLING FACILITY - VENTILATION CONFINEMENT ZONING ANALYSIS

    International Nuclear Information System (INIS)

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

  11. CANISTER HANDLING FACILITY - VENTILATION AIR CALCULATION

    International Nuclear Information System (INIS)

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

  12. COMSOL MULTIPHYSICS MODEL FOR DWPF CANISTER FILLING

    Energy Technology Data Exchange (ETDEWEB)

    Kesterson, M.

    2011-03-31

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

  13. 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

    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...

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

    DEFF Research Database (Denmark)

    Andreasen, A.

    2005-01-01

    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 andthermodynamics the Materials Research...

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

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Furbo, Simon

    point of 58°C and a heat of fusion capacity of 265 kJ/kg. The added xanthane rubber (approx. 2 weight-%) makes the sodium acetate super-cool in a stable way. The starting point for the investigations is an ideal heat storage with perfect heat transfer between charge/discharge fluid and PCM as well......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...

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

    International Nuclear Information System (INIS)

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

  17. 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

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

    International Nuclear Information System (INIS)

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

  19. 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

  20. Stress corrosion cracking of copper canisters

    International Nuclear Information System (INIS)

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

  1. POD (Probability of Detection) evaluation of NDT techniques for Cu-canisters for risk assessment of nuclear waste encapsulation

    International Nuclear Information System (INIS)

    In order to handle long living radioactive waste Sweden is planning to build a deep repository that requires no monitoring by future generations. The spent nuclear fuel will be encapsulated in copper canisters consisting of a graphite cast iron insert shielded by an outer 30-50 mm thick copper cylinder for corrosion and radiation protection. The cast iron insert provides the necessary strength and shielding of radiation. 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 at the canister lab in Oskarshamn, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). Mid 2005 SKB decided that FSW is the preferred sealing technique. A subpart of the final risk assessment for the deep repository is to determine the risk of premature canister leak caused by discontinuities in the insert, in the sealing weld or elsewhere in the copper shielding. Therefore the quality of the production processes and the reliability of the NDT system must be satisfactorily determined and combined to derive assumptions regarding the frequency of undetected production discontinuities in relation to the acceptance criteria for the ensemble of canisters. The reliability of the NDT systems can be derived from POD curves which are investigated for X-ray and ultrasonic techniques applied by SKB. The POD evaluation was carried out by BAM in a joint project for SKB and is evaluated within the common ''a versus a'' approach according to the MIL1823 and some extensions due to the more complex flaw situations in the canisters compared to the original aerospace applications. (orig.)

  2. Storage of nuclear waste in long boreholes

    International Nuclear Information System (INIS)

    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)

  3. Moisture probe using neutron moderation for PuO2 canister inspection

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Shaft shock absorber for a spent fuel canister

    International Nuclear Information System (INIS)

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

  7. 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.

  8. 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 p...... potential: comfort, capacity, efficiency, shifting. •Test of simple control strategies on the elspot price from 2009 in Denmark....

  9. 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

    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......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...

  10. 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.

  11. Pitting corrosion on a copper canister

    International Nuclear Information System (INIS)

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

  12. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    D.T. Dexheimer

    2004-02-27

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

  13. Multi-Canister overpack pressure testing

    International Nuclear Information System (INIS)

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

  14. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    , the MCO is ''N'' stamped as a 450 pounds per square inch (design pressure) vessel. To ensure the process consistently achieves the required weld penetration, a series of developmental tests was performed to identify an optimum and robust set of welding parameters. Testing included test welds made on plate mockups and then actual MCO mockups. With the primary welding parameters (welding current and travel speed) established, a simple two-factor, two-level, factorial experiment with replication at high and low heat input conditions was conducted. Evaluation of the results included weld photomicrographs, which helped establish process range limits for these parameters broad enough to cover typical equipment and measurement variations and provide additional operating margin. To date, over 316 MCOs have been loaded, dried, and transported to the Canister Storage Building (CSB), where the welding is done. Of those, 161 MCOs have received final welded closure and ''N'' stamps. All cover cap final closure welds have met specified requirements without incident

  17. Numerical study of thin layer ring on improving the ice formation of building thermal storage system

    International Nuclear Information System (INIS)

    Ice thermal storage systems have been widely used in HVAC and R systems for improving energy efficiency and reducing energy costs around the world. In this paper, a numerical model is developed to simulate the ice formation in a typical ice thermal storage system. The first study is to investigate the effect of a cooled cylinder placed in a rectangular space filled with water on the ice formation process. The validated numerical model can predict temperature distribution associated with liquid fraction during the process. Based on the result obtained from the first study, further research is focused on the novel structure of thin layer ring. The computational solutions can demonstrate that the thin layer ring structure can successfully increase an ice generated area and shorten the ice formation period in a typical ice thermal storage system. Finally, a parametric study was carried out to investigate the effect of material, thickness, and arrangement of thin layer ring. It predicted that the heat transfer performance of the thin layer ring is dependent on its material, thickness, and arrangement. Ice formation with novel thin layer ring can be improved by increasing the thermal conductivity of a material. A copper ring has the best performance among aluminum, stainless steel, magnesium alloy. The results show that the ice formation rate can be increased by increasing the thickness of the ring from 0.25 mm to 1 mm, while slowed by increasing from 1 mm to 2 mm and has the best performance with 3 mm ring in this study. Finally, the staggered arrangement of ring shows the best results of the ice formation compared to one parallel and two parallel cases. - Highlights: •A thin layer ring structure is studied systematically to enhance ice formation. •Increasing thermal conductivity of thin layer ring can increase ice formation rate. •Ice formation rate is also dependent on the thickness of thin layer ring. •Increasing thin layer ring area can increase ice

  18. EVALUATION OF REQUIREMENTS FOR THE DWPF HIGHER CAPACITY CANISTER

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-05

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    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

  1. 200 Area Interim Storage Area Technical Safety Requirements

    International Nuclear Information System (INIS)

    The 200 Area Interim Storage Area Technical Safety Requirements define administrative controls and design features required to ensure safe operation during receipt and storage of canisters containing spent nuclear fuel. This document is based on the 200 Area Interim Storage Area, Annex D, Final Safety Analysis Report which contains information specific to the 200 Area Interim Storage Area

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

    International Nuclear Information System (INIS)

    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)

  3. Design analysis report for the canister

    International Nuclear Information System (INIS)

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

  4. Dry-type radioactive material storage facility

    International Nuclear Information System (INIS)

    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.)

  5. Dry storage of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  7. Design basis for the copper/steel canister

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-02-01

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

  8. Performance assessment of the copper/steel canister

    International Nuclear Information System (INIS)

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

  9. Design basis for the copper/steel canister

    International Nuclear Information System (INIS)

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

  10. Effect of thermal energy storage in energy consumption required for air conditioning system in office building under the African Mediterranean climate

    Directory of Open Access Journals (Sweden)

    Abdulgalil Mohamed M.

    2014-01-01

    Full Text Available In the African Mediterranean countries, cooling demand constitutes a large proportion of total electrical demand for office buildings during peak hours. The thermal energy storage systems can be an alternative method to be utilized to reduce and time shift the electrical load of air conditioning from on-peak to off-peak hours. In this study, the Hourly Analysis Program has been used to estimate the cooling load profile for an office building based in Tripoli weather data conditions. Preliminary study was performed in order to define the most suitable operating strategies of ice thermal storage, including partial (load leveling and demand limiting, full storage and conventional A/C system. Then, the mathematical model of heat transfer for external ice storage would be based on the operating strategy which achieves the lowest energy consumption. Results indicate that the largest rate of energy consumption occurs when the conventional system is applied to the building, while the lowest rate of energy consumption is obtained when the partial storage (demand limiting 60% is applied. Analysis of results shows that the new layer of ice formed on the surface of the existing ice lead to an increase of thermal resistance of heat transfer, which in return decreased cooling capacity.

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

    International Nuclear Information System (INIS)

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

  12. 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

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

    International Nuclear Information System (INIS)

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

  14. Acceptance of canisters of consolidated spent nuclear fuel by the Federal Waste Management System

    International Nuclear Information System (INIS)

    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: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for deliver; and defense and commercial high-level waste packages. This document discusses canister standards and criteria. 12 refs., 7 figs., 28 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-09

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-09

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

  17. Acceptance of canisters of consolidated spent nuclear fuel by the Federal Waste Management System

    Energy Technology Data Exchange (ETDEWEB)

    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: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for deliver; and defense and commercial high-level waste packages. This document discusses canister standards and criteria. 12 refs., 7 figs., 28 tabs.

  18. Description of DWPF reference waste form and canister

    International Nuclear Information System (INIS)

    This document describes the reference waste form and canister for the Defense Waste Processing Facility (DWPF). The facility is planned for location at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1983. The reference canister is fabricated of 24-in.-OD 304L stainless steel pipe with a dished bottom, domed head, and lifting and welding flanges on the head neck. The overall canister length is 9 ft 10 in., with a wall thickness of 3/8-in. (schedule 20 pipe). The canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected to ensure that a filled canister with its shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be generally compatible with preliminary assessments of repository requireiajps. The rabarajca saspa bkri is bkrksilicapa class cojtaining approximately 28 wt % sludge oxides with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains approximately 58% SiO2 and 15% B2O3. This composition results in a low average leachability in the waste form of approximately 5 x 10-9 g/cm2-day based on 137Cs over 365 days in 250C water. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approximately 425 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the Stage 1 and Stage 2 processes. The radionuclide content of the canister is about 150,000 curies, with a radiation level of 2 x 104 rem/hour at 1 cm

  19. Description of DWPF reference waste form and canister

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

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

  20. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    International Nuclear Information System (INIS)

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data

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

    International Nuclear Information System (INIS)

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

  2. 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

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

    International Nuclear Information System (INIS)

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

  4. Multi-Canister overpack internal HEPA filters

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, K.E.

    1998-11-03

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

  5. National waste terminal storage conceptual reference repository description

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  7. 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

  8. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

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

  9. Spent fuel dry storage technology development: thermal evaluation of sealed storage cask containing spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schmitten, P.F.; Wright, J.B.

    1980-08-01

    A PWR spent fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in a instrumented above surface storage cell during December 1978 for thermal testing. Instrumentation provided to measure canister, liner and concrete temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and liner and in three radial positions in the concrete. Temperatures from the SSC test assembly have been recorded throughout the past 16 months. Canister and liner temperatures have reached their peak values of 200{sup 0}F and 140{sup 0}F, respectively. Computer predictions of the transient and steady-state temperatures show good agreement with the test data.

  10. Spent fuel dry storage technology development: thermal evaluation of sealed storage cask containing spent fuel

    International Nuclear Information System (INIS)

    A PWR spent fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in a instrumented above surface storage cell during December 1978 for thermal testing. Instrumentation provided to measure canister, liner and concrete temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and liner and in three radial positions in the concrete. Temperatures from the SSC test assembly have been recorded throughout the past 16 months. Canister and liner temperatures have reached their peak values of 2000F and 1400F, respectively. Computer predictions of the transient and steady-state temperatures show good agreement with the test data

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  13. Evaluation of remote smearing of DWPF canistered waste forms

    International Nuclear Information System (INIS)

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

  14. Containment canister for capturing hazardous waste debris during piping modifications

    Science.gov (United States)

    Dozier, Stanley B.

    2001-07-24

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  16. Copper canister with cast inner component. Amendment to project on Alternative Systems Study (PASS), SKB TR 93-04

    International Nuclear Information System (INIS)

    The Project on Alternative Systems Study, PASS, was described in a report dated october 1992. In the report, the reference repository concept KBS-3 is described together with three other alternatives. In the report several designs for fuel storage canisters are presented. This report describes a recently developed design for the inner component of the composite, steel and copper, canister which is the main alternative in the KBS-3-model. The new design will be manufactured by casting. A cast insert with inner walls eliminates the need for a stabilizing filler in the canister and guarantees that the fuel remains sub-critical during sufficient time in the repository. The cast insert is judged, to, in comparison with the steel tube alternative, lead to a considerably simplified process in the encapsulation plant and lower development and investment cost. Positive side effects of the design are that the mechanical strength is improved by a factor 2-3 and that the difficult filling operation is avoided in the encapsulation process. The drawbacks are higher weight and probably higher unit price for the empty canister

  17. System-Level Logistics for Dual Purpose Canister Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kalinina, Elena A.

    2014-06-03

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

  18. Development of Thermal Analysis Capability of Dry Storage Cask for Spend Fuel Interim Storage

    International Nuclear Information System (INIS)

    As most of the nuclear power plants, on-site spent fuel pools (SFP) of Taiwan's plants were not originally designed with a storage capacity for all the spent fuel generated over the operating life by their reactors. For interim spent fuel storage, dry casks are one of the most reliable measures to on-site store over-filled assemblies from SFPs. The NUHOMSR-52B System consisting of a canister stored horizontally in a concrete module is selected for thermal evaluation in this paper. The performance of each cask in criticality, radioactive, material and thermal needs to be carefully addressed to ensure its enduring safety. Regarding the thermal features of dry storage casks, three different kinds of heat transfer mechanisms are involved, which include natural convection heat transfer outside and/or inside the canister, radiation heat transfer inside and outside the canister, and conduction heat transfer inside the canister. To analyze the thermal performance of dry storage casks, RELAP5-3D is adopted to calculate the natural air convection and radiation heat transfer outside the canister to the ambient environment, and ANSYS is applied to calculate the internal conduction and radiation heat transfer. During coupling iteration between codes, the heat energy across the canister wall needs to be conserved, and the inner wall temperature of the canister needs to be converged. By the coupling of RELAP5-3D and ANSYS, the temperature distribution within each fuel assembly inside canisters can be calculated and the peaking cladding temperature can be identified. (authors)

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-12-01

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

  1. Technical note 4. Corrosion of copper canister

    International Nuclear Information System (INIS)

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

  2. Multi-canister overpack operations and maintenance manual

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Routt, K.R.

    2001-09-17

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

  4. 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

  5. Annual collection and storage of solar energy for the heating of buildings, report No. 1. Progress report, May--November 1976. [Underground pool of water

    Energy Technology Data Exchange (ETDEWEB)

    Beard, J. T.; Dickey, J. W.; Iachetta, F. A.; Lilleleht, L. U.

    1977-01-01

    A new system for the annual collection and storage of solar heated water for heating of buildings is under development at the University of Virginia. The system is composed of an energy storage sub-system which stores hot water in an underground pool and of a solar collector sub-system which acts not only to collect solar energy throughout the year but also to limit the evaporative and convective heat losses from the storage system. During the summer of 1976, a storage sub-system was constructed using the initial design specifications. A structural failure of that storage pool occurred in August resulting from a leak in the pool liner which caused a failure of the pool structure. A revised design of the storage pool sub-system has been implemented and construction was completed in November, 1976. The collector sub-system has been designed and constructed. Collector operation began in February 1977. A vertical reflector on the north edge of the collector was added in March 1977. Future research will include initial total system operation, performance evaluation, and analytical modeling.

  6. Performance of the SKB copper/steel canister

    International Nuclear Information System (INIS)

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

  7. Canister design for deep borehole disposal of nuclear waste

    OpenAIRE

    Hoag, Christopher Ian.

    2006-01-01

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

  8. Drop tests of the Three Mile Island knockout canister

    International Nuclear Information System (INIS)

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

  9. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

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

  10. Corrosion evaluation of fuel canister crusher rigging

    International Nuclear Information System (INIS)

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

  11. Recent development of spent fuel management in the Central Storage Facility for Spent Fuel (CLAB) in Sweden

    International Nuclear Information System (INIS)

    The Swedish nuclear waste management programme was presented at the Advisory Group Meeting held 15-18 March 1988. As there have been very small changes since then in the main strategy this paper is focusing the ongoing work to expand the capacity of the Central Storage Facility for Spent Fuel, CLAB, from 3.000 tonnes of uranium (tU) to 5.000 tU within the existing storage pools. This expansion makes it possible to postpone the investment in new storage pools in a second rock cavern by 6-8 years which gives a considerable economic advantage. Preliminary studies of several methods to realize the expansion led to a decision to use storage canisters similar to the existing ones. The new BWR canister will be designed to hold 25 fuel assemblies instead of 16 and the new PWR canister 9 assemblies instead of 5. The main problem is to maintain a sufficient margin against criticality with the new and denser fuel pattern. Two methods to achieve this have been more closely investigated: Credit for the burnup of the fuel; Neutron absorbing material in the canisters for reactivity control. A calculational work has been performed in order to analyze the margins which must be applied in the first case. One of the more important factors proved to be the margin required to cover the influence of axial profiles of burnup and Pu-build up for BWR fuel. Together with other necessary reactivity margins the total penalty which has to be applied amounted to about 12 reactivity percent units. With the requirement of a final Keff value smaller than or equal to 0,95 this implied that a far to great proportion of the fuel arriving at CLAB would fall outside the acceptable region defined by initial enrichment and actual burnup. The method of taking full credit for fuel burnup therefore had to be abandoned in favour of the other method. A preliminary analysis has been performed involving different degrees of absorber material in the canister and the fact that burnable absorbers in the fuel

  12. Atomic storage

    CERN Multimedia

    Ricadela, A

    2003-01-01

    IBM is supplying CERN, the European Organization for Nuclear Research, with its Storage Tank file system virtualization software, 20 terabytes of storage capacity, and services under a three-year deal to build computer systems that will support the Large Hadron Collider accelerator (1 paragraph).

  13. Concept of BN-350 reactor spent fuel handling during its storage after shutdown

    International Nuclear Information System (INIS)

    Full text: According to the Kazakhstan Government Decree (456; Apr 22, 1999), the fast BN-350 reactor has been shutdown in 1999 and the plan of its decommissioning has been started. By a decision of the Kazakhstan Government is defined that its spent fuel must be placed into 50 year's long-term safe storage with following dismantling and final disposal. This plan most important part is the concept of the spent fuel handling during its storage after shutdown. Next main stages of this concept are discussed here: discharge of the fuel, spent fuel packaging into special canisters, temporary storage of these canisters in the rector pool, to choose the site for long-term spent fuel storage and transport cask, making a choice of safety technique for canisters dry storage during 50 years. At first stage the fuel has been removed from the core and placing into reactor pool. The second stage lasted practically about two years. During this period all spent fuel has been packaged into special sealed canisters filled with inert gas. On next stage the site outside of the reactor one for long-term storage of these canisters and the way to ship them into the casks were chosen. The selected site is placed within the territory of former Semipalatinsk nuclear site. But the casks for spent fuel ought to be shipping as by rail and trailer vehicles too. Until its shipping will be started the canisters have been temporary stored under the water into BN-350 reactor pool. The following step is to define the transport cask design and the way to store the canisters at the site during long-term period about 50 years. There are two projects, which have been under consideration. The first is to use single-canister iron cask designed for shipping only. In this case the canisters with spent fuel ought to be transporting to the site into these casks and then transshipped into special 'silo'-type storage. Each canister has to be placed into individual near surface silo. The goal of this way is to

  14. 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

  15. 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)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  18. 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)

    Proposed action is to construct at BNL a 5,600-ft2 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

  19. 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. Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

    Proposed action is to construct at BNL a 5,600-ft{sup 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.

  20. 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

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

    Proposed action is to construct at BNL a 5,600-ft[sup 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.

  1. Mechanical Integrity of Canisters Using a Fracture Mechanics Approach

    International Nuclear Information System (INIS)

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

  2. 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

  3. Design, production and initial state of the canister

    International Nuclear Information System (INIS)

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

  4. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-11-15

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

  5. 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.)

  6. MACSTOR trademark: Dry spent fuel storage for the nuclear power industry

    International Nuclear Information System (INIS)

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL's successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection

  7. Decontamination of DWPF canisters by glass frit blasting

    International Nuclear Information System (INIS)

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

  8. Radiation-field mapping of insect irradiation canisters

    International Nuclear Information System (INIS)

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

  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. Lesson Learned from Drop and Tip-Over Test of a Dry Storage Cask System

    International Nuclear Information System (INIS)

    To study an appropriate storage system with the consideration of a nuclear power plant situation, a status evaluation of the technology and a feasibility study has been performed in Korea. As a part of doing these, a dry spent fuel storage cask has been developed. The dry storage cask under development consists of a cask body, a canister, and an in-canister structure. 24 fuel assemblies are stored in canister structure. The spacer disks in this in-canister structure are designed to dissipate the heat from the baskets and to provide a lateral support to the baskets. The support rods keep the spacer disks at an even interval. To assess a structural integrity after a hypothetical accident condition of this dry storage system, a 1/4 scale model is fabricated for the drop test. Drop tests of this test model were performed and the test results were also discussed

  11. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    International Nuclear Information System (INIS)

    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

  12. 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

  13. Chemical stability of copper-canisters in deep repository

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Effects of brine migration on waste storage systems. Final report

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Kjell [Matsafe AB, Stockholm (Sweden)

    2006-12-15

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

  17. 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)

    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

  18. Spent fuel dry storage technology development: electrically heated drywell storage test (3kW operation)

    International Nuclear Information System (INIS)

    An electrically heated drywell storage cell test has been in operation since March 1978 at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data obtained at electric heater power output of 3.0 kW and compares the data with that for heater power outputs of 1.0 kW and 2.0 kW. The simulated drywell storage cell consists of a stainless steel canister (representative of the spent fuel canisters being tested at E-MAD) containing an electrical heater assembly, a concrete-filled shield plug to which the canister is attached, and a carbon steel liner that encloses the canister and shield plug. The entire test drywell is grouted into a hole drilled in the soil adjacent to E-MAD. Temperature instrumentation is provided on the exterior of the canister and liner, in the grout around the liner, and at six radial locations in the soil surrounding the drywell. Peak measured canister and liner temperatures are 7850F and 7470F, respectively. Previous testing showed peak measured canister and liner temperatures of 2760F and 2320F for 1.0 kW and 5060F and 4580F for 2.0 kW, respectively. A previously developed computer model was utilized to predict the thermal response of the test configuration. Computer predictions of the transient and steady-state temperatures of the drywell components and surrounding soil are presented and are compared with the test data

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

    International Nuclear Information System (INIS)

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

  20. Internals removal and storage options for an actual annealing

    International Nuclear Information System (INIS)

    The objective of this paper was an overview of the options for storage of the reactor internals during an actual annealing process. Physical support of the internals, adequate shielding, and seals to protect the vessel from water leakage were considered in this review. Options considered were: (1) Canister approach, (2) Dam approach, and (3) Dry storage shielded cask approach

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

    International Nuclear Information System (INIS)

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

  2. 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.

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

    International Nuclear Information System (INIS)

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

  4. Multi-Canister overpack dual pressure rating; TOPICAL

    International Nuclear Information System (INIS)

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

  5. Effects of glacial meltwater on corrosion of copper canisters

    International Nuclear Information System (INIS)

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

  6. The Universal Canister Strategy in Spent Fuel Reprocessing: UC-C a Real Industrial Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Thomasson, J.; Barithel, S.; Cocaud, A.; Derycke, P.; Pierre, P.

    2003-02-25

    In commercial nuclear activities, spent fuel back end management is a key issue for nuclear countries as spent fuel represent most of national civil nuclear waste legacy. Ensuring public safety and protection of the environment, now and in the future has been and still remains a major commitment, it is still the subject of thorough development efforts and active public debates. Considerable benefits can be obtained from the Universal Canister strategy as implemented in France in spent fuel treatment and waste conditioning based on reprocessing. COGEMA developed sophisticated waste conditioning processes to simplify High Level and Long Lived Intermediate Level Waste storage and final disposal. Main benefits are: waste stabilization by immobilization and encapsulation; ultimate waste toxicity reduction; drastic ultimate waste volume reduction; and ultimate waste packages standardization.

  7. Work plan and health and safety plan for Building 3019B underground storage tank at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Burman, S.N.; Brown, K.S.; Landguth, D.C.

    1992-08-01

    As part of the Underground Storage Tank Program at the Department of Energy's Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, this Health and Safety Plan has been developed for removal of the 110-gal leaded fuel underground storage tank (UST) located in the Building 3019B area at ORNL This Health and Safety Plan was developed by the Measurement Applications and Development Group of the Health and Safety Research Division at ORNL The major components of the plan follow: (1) A project description that gives the scope and objectives of the 110-gal tank removal project and assigns responsibilities, in addition to providing emergency information for situations occurring during field operations; (2) a health and safety plan in Sect. 15 for the Building 3019B UST activities, which describes general site hazards and particular hazards associated with specific tasks, personnel protection requirements and mandatory safety procedures; and (3) discussion of the proper form completion and reporting requirements during removal of the UST. This document addresses Occupational Safety and Health Administration (OSHA) requirements in 29 CFR 1910.120 with respect to all aspects of health and safety involved in a UST removal. In addition, the plan follows the Environmental Protection Agency (EPA) QAMS 005/80 (1980) format with the inclusion of the health and safety section (Sect. 15).

  8. Work plan and health and safety plan for Building 3019B underground storage tank at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Burman, S.N.; Brown, K.S.; Landguth, D.C.

    1992-08-01

    As part of the Underground Storage Tank Program at the Department of Energy`s Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, this Health and Safety Plan has been developed for removal of the 110-gal leaded fuel underground storage tank (UST) located in the Building 3019B area at ORNL This Health and Safety Plan was developed by the Measurement Applications and Development Group of the Health and Safety Research Division at ORNL The major components of the plan follow: (1) A project description that gives the scope and objectives of the 110-gal tank removal project and assigns responsibilities, in addition to providing emergency information for situations occurring during field operations; (2) a health and safety plan in Sect. 15 for the Building 3019B UST activities, which describes general site hazards and particular hazards associated with specific tasks, personnel protection requirements and mandatory safety procedures; and (3) discussion of the proper form completion and reporting requirements during removal of the UST. This document addresses Occupational Safety and Health Administration (OSHA) requirements in 29 CFR 1910.120 with respect to all aspects of health and safety involved in a UST removal. In addition, the plan follows the Environmental Protection Agency (EPA) QAMS 005/80 (1980) format with the inclusion of the health and safety section (Sect. 15).

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

    International Nuclear Information System (INIS)

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

  10. 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.

  11. 42 CFR 84.1153 - Dust, fume, mist, and smoke tests; canister bench tests; gas masks canisters containing filters...

    Science.gov (United States)

    2010-10-01

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

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

    International Nuclear Information System (INIS)

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

  13. A passive air-cooled dry storage facility for vitrified high-level wastes

    International Nuclear Information System (INIS)

    A conceptual design of air-cooled dry storage vault facility for vitrified high-level waste (HLW) canisters is developed for a site in northern Japan. The facility is designed for the reception and unloading of shielded seagoing transportation casks of vitrified HLW canisters, for the inspection of these canisters, and for their temporary storage for a period of up to 50 years. The waste is to be at least 9 years old when received, and the facility will be capable of storing up to 2,500 canisters. This paper provides a conceptual design to identify construction requirements, materials, and space requirements that are unique to the vitrified HLW storage facility. It also identifies the types of special systems and equipment needed in such a facility

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

    International Nuclear Information System (INIS)

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

  15. Evaluation of closure alternatives for the Building 3001 Storage Canal at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    The Bldg. 3001 Storage Canal at ORNL is leaking approximately 400 gal of water per day. This report presents the Bechtel National Inc. (BNI) Team's evaluation of plans and presents recommendations for interim closure alternatives to stop the release of radionuclides and potential release of heavy metals into the environment. This is a conceptual evaluation and does not include detailed engineering of physical mitigation methods. The alternatives address only interim closure measures and not final decommissioning of the canal

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A.J.

    1995-12-31

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

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

    International Nuclear Information System (INIS)

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

  19. HABOG: One building for all high level waste and spent fuel in the Netherlands. The first year of experience

    International Nuclear Information System (INIS)

    COVRA N.V. operates a site for the treatment and storage of all types of radioactive waste in the Netherlands. In 2003 the facility for storage of high level reprocessing waste and spent fuel from research reactors came into operation. The building is a piece of art. In a technical sense, as it was designed to withstand events such as earthquakes, flooding and aircraft crashes; but also in a literal sense, its provocative design was meant to evoke a discussion about the public aspects of radioactive waste management. Construction took place from 1998 until 2002 and active operation at the end of 2003. The technical experience is positive. Nine MTR2 containers with spent fuel were treated and stored and 28 canisters with vitrified reprocessing waste were stored. The experience with the communicative function is positive also. Since the opening of the facility, the amount of visitors to the COVRA site more than doubled. (author)

  20. Facility site check report transportation safeguards divsision (TSD) underground storage tanks 2334-U and 2335-U at Building 9714

    International Nuclear Information System (INIS)

    This document presents an overview of the underground storage tank (UST)-related events that have taken place at the Transportation Safeguards Division (TSD) Facility (Facility ID 0-730168). The TSD facility is managed by Lockheed Martin Energy Systems, Inc. (LMES) for the U.S. Department of Energy (DOE), and is used to maintain and fuel specialty fleet vehicles. The facility is located approximately one mile east of the K-25 site at the intersection of Blair Road and the Oak Ridge Turnpike (Hwy 58). The location of the USTs at the TSD facility are illustrated

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

    International Nuclear Information System (INIS)

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

  2. Testing and evaluating storage technology to build a distributed Tier1 for SuperB in Italy

    International Nuclear Information System (INIS)

    The SuperB asymmetric energy e+e−- collider and detector to be built at the newly founded Nicola Cabibbo Lab will provide a uniquely sensitive probe of New Physics in the flavor sector of the Standard Model. Studying minute effects in the heavy quark and heavy lepton sectors requires a data sample of 75 ab−-1 and a luminosity target of 1036 cm−-2 s−-1. This luminosity translate in the requirement of storing more than 50 PByte of additional data each year, making SuperB an interesting challenge to the data management infrastructure, both at site level as at Wide Area Network level. A new Tier1, distributed among 3 or 4 sites in the south of Italy, is planned as part of the SuperB computing infrastructure. Data storage is a relevant topic whose development affects the way to configure and setup storage infrastructure both in local computing cluster and in a distributed paradigm. In this work we report the test on the software for data distribution and data replica focusing on the experiences made with Hadoop and GlusterFS.

  3. Testing and evaluating storage technology to build a distributed Tier1 for SuperB in Italy

    Science.gov (United States)

    Pardi, S.; Fella, A.; Bianchi, F.; Ciaschini, V.; Corvo, M.; Delprete, D.; Di Simone, A.; Donvito, G.; Giacomini, F.; Gianoli, A.; Longo, S.; Luitz, S.; Luppi, E.; Manzali, M.; Perez, A.; Rama, M.; Russo, G.; Santeramo, B.; Stroili, R.; Tomassetti, L.

    2012-12-01

    The SuperB asymmetric energy e+e-- collider and detector to be built at the newly founded Nicola Cabibbo Lab will provide a uniquely sensitive probe of New Physics in the flavor sector of the Standard Model. Studying minute effects in the heavy quark and heavy lepton sectors requires a data sample of 75 ab--1 and a luminosity target of 1036 cm--2 s--1. This luminosity translate in the requirement of storing more than 50 PByte of additional data each year, making SuperB an interesting challenge to the data management infrastructure, both at site level as at Wide Area Network level. A new Tier1, distributed among 3 or 4 sites in the south of Italy, is planned as part of the SuperB computing infrastructure. Data storage is a relevant topic whose development affects the way to configure and setup storage infrastructure both in local computing cluster and in a distributed paradigm. In this work we report the test on the software for data distribution and data replica focusing on the experiences made with Hadoop and GlusterFS.

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

    Science.gov (United States)

    Levine, Howard G.; Caron, Allison

    2016-01-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

  6. Analysis of probability of defects in the disposal canisters

    International Nuclear Information System (INIS)

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

  7. CLASSIFICATION OF THE MGR CANISTERED SNF DISPOSAL CONTAINER SYSTEM

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-02-01

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

  11. Test report for the Sample Transfer Canister system

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, B.D.

    1998-03-04

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

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

    International Nuclear Information System (INIS)

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

  13. Thermal Dimensioning of SiC Canister Applied A-KRS

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  14. Thermal Dimensioning of SiC Canister Applied A-KRS

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, C.G

    1998-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. 32-Week Holding-Time Study of SUMMA Polished Canisters and Triple Sorbent Traps Used To Sample Organic Constituents in Radioactive Waste Tank Vapor Headspace

    International Nuclear Information System (INIS)

    Two sampling methods[SUMMA polished canisters and triple sorbent traps (TSTs)] were compared for long-term storage of trace organic vapor samples collected from the headspaces of high-level radioactive waste tanks at the U.S. Department of Energy's Hanford Site in Washington State. Because safety, quality assurance, radiological controls, the long-term stability of the sampling media during storage needed to be addressed. Samples were analyzed with a gas chromatograph/mass spectrometer (GC/MS) using cryogenic reconcentration or thermal desorption sample introduction techniques. SUMMA canister samples were also analyzed for total non-methane organic compounds (TNMOC) by GC/flame ionization detector (FID) using EPA Compendium Method TO-12 . To verify the long-term stability of the sampling media, multiple samples were collected in parallel from a typical passively ventilated radioactive waste tank known to contain moderately high concentrations of both polar and nonpolar organic compounds. Analyses for organic analytes and TNMOC were conducted at increasing intervals over a 32-week period to determine whether any systematic degradation of sample integrity occurred. Analytes collected in the SUMMA polished canisters generally showed good stability over the full 32 weeks with recoveries at the 80% level or better for all compounds studied. The TST data showed some loss (50-80% recovery) for a few high-volatility compounds even in the refrigerated samples; losses for unrefrigerated samples were far more pronounced with recoveries as low as 20% observed in a few cases

  1. Alternatives for water basin spent fuel storage using pin storage

    International Nuclear Information System (INIS)

    The densest tolerable form for storing spent nuclear fuel is storage of only the fuel rods. This eliminates the space between the fuel rods and frees the hardware to be treated as non-fuel waste. The storage density can be as much as 1.07 MTU/ft2 when racks are used that just satisfy the criticality and thermal limitations. One of the major advantages of pin storage is that it is compatible with existing racks; however, this reduces the storage density to 0.69 MTU/ft2. Even this is a substantial increase over the 0.39 MTU/ft2 that is achievable with current high capacity stainless steel racks which have been selected as the bases for comparison. Disassembly requires extensive operation on the fuel assembly to remove the upper end fitting and to extract the fuel rods from the assembly skeleton. These operations will be performed with the aid of an elevator to raise the assembly where each fuel rod is grappled. Lowering the elevator will free the fuel rod for transfer to the storage canister. A storage savings of $1510 per MTU can be realized if the pin storage concept is incorporated at a new away-from-reactor facility. The storage cost ranges from $3340 to $7820 per MTU of fuel stored with the lower cost applying to storage at an existing away-from-reactor storage facility and the higher cost applying to at-reactor storage

  2. Validation of the FLUENT CFD Computer Program by Thermal Testing of a Full Scale Double-Walled Prototype Canister for Storing Chernobyl Fuel

    International Nuclear Information System (INIS)

    To provide a high degree of confidence in the results predicted by the FLUENT CFD computer code for safety evaluation of storing Chernobyl spent nuclear fuel (SNF) in double-walled canisters (DWC) a full scale prototype DWC was manufactured and tested at the Holtec Manufacturing Division in Turtle Creek, PA. The DWC was instrumented and fuel heat simulated by inserting electrically heated rods in storage cells under two extreme heat distribution scenarios: Core heated test wherein the heat is applied to the innermost storage cells and Peripherally heated test wherein the heat is applied to the outermost storage cells. The heater tubes, storage cells, DWC shell and lid were instrumented to measure and record temperatures during the testing. To validate the FLUENT CFD code the thermal tests were simulated on FLUENT by constructing geometrically accurate 3D model of the DWC with all internals significant to mimic the thermal-hydraulic state in the DWC. These included heated rods, fuel tubes, support plates and the DWC shell. The test measurements and FLUENT simulations were evaluated and the predictability of the FLUENT CFD code for safety evaluation of fuel storage in double-walled canisters confirmed. (authors)

  3. Novel approach for decentralized energy supply and energy storage of tall buildings in Latin America based on renewable energy sources: Case study – Informal vertical community Torre David, Caracas – Venezuela

    International Nuclear Information System (INIS)

    This paper analyzes the concept of a decentralized power system based on wind energy and a pumped hydro storage system in a tall building. The system reacts to the current paradigm of power outage in Latin American countries caused by infrastructure limitations and climate change, while it fosters the penetration of renewable energy sources (RES) for a more diversified and secure electricity supply. An explicit methodology describes the assessment of technical, operational and economic potentials in a specific urban setting in Caracas/Venezuela. The suitability, applicability and the impacts generated by such power system are furthermore discussed at economic, social and technical level. - Highlights: ► We have modeled an innovative pico pumped hydro-storage system and wind power system for tall buildings. ► We conducted technical, economic and social analysis on these energy supply and storage alternatives. ► The energy storage system can achieve efficiencies within 30% and 35%. ► The energy storage is realistic and economic sensible in comparison to other solutions. ► The impacts of such a system in the current living conditions and safety issues of the building are minimum

  4. Reactor building

    International Nuclear Information System (INIS)

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  5. Expertise concerning the request by the ZWILAG Intermediate Storage Facility Wuerenlingen AG for granting of a licence for the building and operation of the Central Intermediate Storage Facility for radioactive wastes

    International Nuclear Information System (INIS)

    On July 15, 1993, the Intermediate Storage Facility Wuerenlingen AG (ZWILAG) submitted a request to the Swiss Federal Council for granting of a license for the construction and operation of a central intermediate storage facility for radioactive wastes. The project foresees intermediate storage halls as well as conditioning and incineration installations. The Federal Agency for the Safety of Nuclear Installations (HSK) has to examine the project from the point of view of nuclear safety. The present report presents the results of this examination. Different waste types have to be treated in ZWILAG: spent fuel assemblies from Swiss nuclear power plants (KKWs); vitrified, highly radioactive wastes from reprocessing; intermediate and low-level radioactive wastes from KKWs and from reprocessing; wastes from the dismantling of nuclear installations; wastes from medicine, industry and research. The wastes are partitioned into three categories: high-level (HAA) radioactive wastes containing, amongst others, α-active nuclides, intermediate-level (MAA) radioactive wastes and low-level (SAA) radioactive wastes. The projected installation consists of three repository halls for each waste category, a hot cell, a conditioning plant and an incineration and melting installation. The HAA repository can accept 200 transport and storage containers with vitrified high-level wastes or spent fuel assemblies. The expected radioactivity amounts to 1020 Bq, including 1018 Bq of α-active nuclides. The thermal power produced by decay is released to the environment by natural circulation of air. The ventilation system is designed for a maximum power of 5.8 MW. Severe conditions are imposed to the containers as far as tightness and shielding against radiation is concerned. In the repository for MAA wastes the maximum radioactivity is 1018 Bq with 1015 Bq of α-active nuclides. The maximum thermal power of 250 kW is removed by forced air cooling. Because of the high level of radiation the

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Introductory remarks to the Workshop on storage with surveillance versus immediate decommissioning for nuclear reactor components and buildings

    International Nuclear Information System (INIS)

    This paper introduces the dicotomy of immediate versus delayed dismantling to the Committee but its theme is to remind members of the third alternative of no dismantling at all. Two necessary conditions for immediate dismantling are the availability of waste disposal sites and the availability of the necessary funds. A strong financial argument for delayed dismantling is based upon the theory of long term discounting. This theoretical premise however, neglects the said history of social, fiscal and corporate instability over periods of the order fo fifty years. Those arguments for either immediate or delayed dismantling associated with radioactive exposure are not conclusive because the total activity does not decay significantly after the first thirty years, and radiation field from significantly decayed Co60 sources can still be sufficient to necessitate remote handling techniques. The advantages of avoiding the dismantling option altogether should be thoroughly investigated because the risk to the public entailed by leaving a reliable containment building intact can, in most cases be shown to be negligible in relation to other daily risks. The use of the funds so released can be better utilized in reducing the level of other more urgent risks

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

    International Nuclear Information System (INIS)

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

  10. Experimental study of the application of intermittently operated SEHRAC (storage-enhanced heat recovery room air-conditioner) in residential buildings in Hong Kong

    International Nuclear Information System (INIS)

    Effectiveness of SEHRAC (storage-enhanced heat recovery room air-conditioner) for water heating in residential buildings in Hong Kong and elsewhere has been confirmed in previous studies. However, given these studies assumed a theoretical maximum recoverable heat, whether its use is still energy effective in practice, in particular under intermittent operation, is of concern. Intermittent operation of the SEHRAC can lead to significant fluctuations in operating conditions. Adding that capillary tube is often used as the expansion device to magnify the fluctuations, whether SEHRAC can still operate satisfactorily despite the fluctuations is another concern. To address these concerns, a prototype which can be switched between the combined CH (cooling and heating) mode and the CC (conventional cooling) mode was set-up for laboratory experiments. The results showed that the water heating objective can be achieved. The operating parameters also confirmed the satisfactory operation of SEHRAC. Energy performance of the CH mode was found better than the CC mode. A prediction model was developed for evaluating the use of SEHRAC. On wider application of SEHRAC, energy use of the residential sector in Hong Kong can be reduced by 9.1%. The experimental details described in this study would become an experiment protocol to enhance future research in this area. - Highlights: • Practical use of SEHRAC (storage-enhanced heat recovery room air-conditioner) for free water heating was investigated. • Investigations were based on laboratory experiments that matched with practical situations. • Experimental results confirmed the effective operation of SEHRAC in practical situations. • Potential water heating energy saving on wider application of SEHRAC was estimated to be 9.1%. • The prototype designed and set-up for this study would become an experiment protocol to enhance future research in this area

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-02-01

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

  12. 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

    2009-03-10

    Berkeley Lab has for several years been developing methods for selection of optimal microgrid systems, especially for commercial building applications, and applying these methods in the Distributed Energy Resources Customer Adoption Model (DER-CAM). 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 electricity storage, and (3) to make an initial effort towards adding consideration of power quality and reliability (PQR) to the capabilities of DER-CAM. All of these objectives have been pursued via analysis of the attractiveness of a Consortium for Electric Reliability Technology Solutions (CERTS) Microgrid consisting of multiple nameplate 100 kW Tecogen Premium Power Modules (CM-100). This unit consists of an asynchronous inverter-based variable speed internal combustion engine genset with combined heat and power (CHP) and power surge capability. The essence of CERTS Microgrid technology is that smarts added to the on-board power electronics of any microgrid device enables stable and safe islanded operation without the need for complex fast supervisory controls. This approach allows plug and play development of a microgrid that can potentially provide high PQR with a minimum of specialized site-specific engineering. A notable feature of the CM-100 is its time-limited surge rating of 125 kW, and DER-CAM capability to model this feature was also a necessary model enhancement.

  13. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

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

    International Nuclear Information System (INIS)

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

  15. The P6 truss moves to a payload transport canister

    Science.gov (United States)

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

  17. Heat transfer study on dry vault storage system

    International Nuclear Information System (INIS)

    IHI has been studying the dry vault storage system based on the experience of the vitrified products storage facility. Maximum allowable temperature of fuel cladding was decided by creep strain criteria for long term dry storage environment to avoid cladding degradation. It was necessary to establish the evaluation method of heat transfer inside and outside the fuel loaded canisters for the design of storage facility. Therefore, the experimental and analytical studies of heat transfer of dry vault storage system were carried out using the experimental apparatus and the analysis program based on finite element method. (author)

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

    International Nuclear Information System (INIS)

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

  19. Commercial solutions [for dry spent fuel storage casks

    International Nuclear Information System (INIS)

    In the aftermath of the termination of the DOE's MPC (Multi-Purpose Canister) programme, commercial suppliers are coming forward with new or updated systems to meet utility needs. Leading vendors describe the advantages of their systems for dry spent fuel storage and transport. (Author)

  20. Thermal energy storage application areas

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-01

    The use of thermal energy storage in the areas of building heating and cooling, recovery of industrial process and waste heat, solar power generation, and off-peak energy storage and load management in electric utilities is reviewed. (TFD)