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Sample records for canister storage building

  1. 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

  2. Canister storage building hazard analysis report

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  3. Canister storage building trade study. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

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

  4. Canister storage building trade study. Final report

    International Nuclear Information System (INIS)

    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 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.

  6. 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

  7. 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

  8. 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.

  9. 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

  10. Canister Storage Building (CSB) Hazard Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    2000-03-16

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

  11. 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.

  12. 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.

  13. 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

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dana, W.P.

    1995-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    TU, K.C.

    1999-10-08

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

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

    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.

  18. 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

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

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

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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.

    2000-06-07

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

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

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

  7. 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.

  8. 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

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-29

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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D.

    1986-04-01

    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.

  13. 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

  14. 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.

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E.

    1978-12-01

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

  17. 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.

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-08

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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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

  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. 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)

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

    International Nuclear Information System (INIS)

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

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  2. 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)

  3. 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)

  4. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    B. Gorpani

    2000-06-23

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

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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)

  19. 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 ...

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Deffenbaugh, M.L.

    1998-05-28

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

  2. 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

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

  4. 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...

  5. 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.

  6. 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...

  7. 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

  8. 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.

  9. 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

  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. 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.

  12. 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

  13. Staging and storage facility feasibility study. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-02-01

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

  14. 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.

  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...... walls of light-weight concrete. The design promise stated that a few years of dehumidification would bring down the moisture contained in the fresh constructions to a level corresponding with the desired interior climate. After this initial stage, the passive control would eliminate all further need...... 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. 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

  17. 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)

  18. 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

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  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. 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.

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  15. 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

    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...

  16. Corrosion resistant storage container for radioactive material

    Science.gov (United States)

    Schweitzer, Donald G.; Davis, Mary S.

    1990-01-01

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

  17. 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.

  18. 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.

  19. 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

    The use of thermal energy storage (TES) in buildings in combination with space heating and/or space cooling has recently received much attention. A variety of TES techniques have developed over the past decades. TES systems can provide short-term storage for peak-load shaving as well as long-term...

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

    Energy Technology Data Exchange (ETDEWEB)

    C.A Kouts

    2006-11-22

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

  1. 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...

  2. 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

  3. 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...

  4. 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

  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. 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

  8. 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

  9. Air quality in low-ventilated museum storage buildings

    DEFF Research Database (Denmark)

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

    2014-01-01

    internally-generated pollutants accumulate, were measured by passive sampling of ozone, nitrogen dioxide, and organic acids. The air exchange rates and the interchange of air between storage rooms were measured by the per-fluorocarbon tracer gas method. Ambient pollutants were reduced in concentration...

  10. 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

  11. 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.

  12. 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

  13. 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

  14. 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

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

  15. 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

  16. 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.

  17. 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...... systems through further improvement of water based storages and in parallel to investigate the potential of using storage designs with phase change materials, PCM. The advantage of phase change materials is that large amounts of energy can be stored without temperature increase when the material is going...... from solid to liquid form (Fig. 1). Keeping the temperature as low as possible is an efficient way to reduce the heat loss from the storage. Furthermore, the PCM storage might be smaller than the equivalent water storage as more energy can be stored per volume. If the PCM further has the possibility...

  18. 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

  19. 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)

  20. 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.)

  1. Canister compatibility with Carlsbad salt

    International Nuclear Information System (INIS)

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

  2. Canister Transfer Facility Criticality Calculations

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Monroe-Rammsy

    2000-10-13

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

  3. Fire Design Study Case of a High-Rise Steel Storage Building

    OpenAIRE

    Franssen, Jean-Marc; Zaharia, Raoul

    2002-01-01

    This paper presents a fire design study case for a high-rise storage rach supporting building. Standardised ISO and natural fire models were considered for the fire action. The structural analysis was carried out by means of the advanced numerical program SAFIR, an FEM software specialised for the thermal and mechanical analysis of structures submitted to the fire. Peer reviewed

  4. 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...

  5. 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

  6. Spent fuel canister docking station

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-15

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

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

    DEFF Research Database (Denmark)

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

    2007-01-01

    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...... as their interrelation. The potential of increasing thermal mass by using phase change materials (PCM) was estimated assuming increased thermal capacity. The results show a significant impact of the heat transfer coefficient on heat storage capacity, especially for thick, thermally heavy elements. The storage capacity...... of a 100 mm thick concrete slab was found to increase with increasing heat transfer coefficients as high as 30 W/m2K. In contrast the heat storage capacity of a thin gypsum plaster board was found to be constant when the heat transfer coefficient exceeded 3 W/m2K. Additionally, the optimal thickness...

  10. 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

  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. 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

  13. 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

  14. 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.

  15. 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...

  16. 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.

  17. 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

  18. Hygrothermal evaluation of a museum storage building based on actual measurements and simulations

    DEFF Research Database (Denmark)

    Christensen, Jørgen Erik; Kollias, Christos Georgios

    2015-01-01

    Museum storage buildings should be able to provide a considerable stable indoor environment in terms of temperature and relative humidity (RH). To obtain such stable conditions with the lowest possible energy consumption, passive air conditioning is one-way solution. In this paper, indoor...... to maintain RH within acceptable levels. Therefore, renewable energy such us excess wind energy during the night can be utilized. [All rights reserved Elsevier]....

  19. Eutectic mixtures of capric acid and lauric acid applied in building wallboards for heat energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Shilei, L.; Neng, Z. [School of Environment Science and Technology, Tianjin University, Tianjin (China); Guohui, F. [Shenyang Jianzhu University, Shenyang (China)

    2006-07-01

    Capric acid (CA) and lauric acid (LA), as phase change materials (PCM), can be applied for energy storage in low temperature. The phase transition temperature and values of latent heat of eutectic mixtures of CA and LA are suitable for being incorporated with building materials to form phase change wallboards used for building energy storage. 120, 240 and 360 accelerated thermal cycle tests were conducted to study the changes in latent heat of fusion and melting temperature of phase change wallboards combined with the eutectic mixtures of CA and LA. Differential scanning calorimetry (DSC) tested the transition temperature and latent heat. The results showed that the melting temperature and latent heat of these phase change wallboards with eutectic mixtures have no obvious variations after repeated 360 thermal cycles, which proved that these phase change wallboards have good thermal stability for melting temperature and variations in latent heat of fusion for long time application. Therefore, they can be used for latent heat storage in the field of building energy conservation. (author)

  20. Eutectic mixtures of capric acid and lauric acid applied in building wallboards for heat energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Lv Shilei; Zhu Neng [Tianjin University (China). School of Environmental Science and Technology; Feng Guohui [Shenyang Jianzhu University, Shenyang (China)

    2006-06-15

    Capric acid (CA) and lauric acid (LA), as phase change materials (PCM), can be applied for energy storage in low temperature. The phase transitions temperature and values of latent heat of eutectic mixtures of CA and LA are suitable for being incorporated with building materials to form phase change wallboards used for building energy storage. 120, 240 and 360 accelerated thermal cycle tests were conducted to study the changes in latent heat of fusion and melting temperature of phase change wallboards combined with the eutectic mixtures of CA and LA. Differential scanning calorimetry (DSC) tested the transition temperature and latent heat. The results showed that the melting temperature and latent heat of these phase change wallboards with eutectic mixtures have not obvious variations after repeated 360 thermal cycles, which proved that these phase change wallboards have good thermal stability for melting temperature and variations in latent heat of fusion for long time application. Therefore, they can be used for latent heat storage in the field of building energy conservation. (author)

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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

  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. 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.

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Andreasen, Anders

    2005-09-01

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

  12. 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

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

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

  15. 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

  16. 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.

  17. 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.

  18. 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.

  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. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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.)

  6. 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.

  7. 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.)

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

    International Nuclear Information System (INIS)

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

  9. 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

  10. 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.

  11. 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.

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

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

  14. 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).

  15. 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

  16. 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

  17. 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.

  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

    to achieve 100% coverage of space heating and domestic hot water in a low energy house in a Danish climate with a solar heating system with 36 m² flat plate solar collector and approximately 10 m³ storage with sodium acetate. A traditional water storage solution aiming at 100% coverage will require a storage...

  19. 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)

  20. 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

  1. 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

    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......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...... 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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-29

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

  3. 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

  4. 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

  5. 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.

  6. 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...

  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. 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.

  9. 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)

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

    DEFF Research Database (Denmark)

    Heller, Alfred

    1996-01-01

    solar collector area of the system, was achieved. Active heating from the sand storage was not observed. The pay-back time for the system can be estimated to be similar to solar heated domestic hot water systems in general. A number of minor improvements on the system could be pointed out.......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...

  11. 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)

  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. 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

  14. 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

  15. 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.

  16. 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,…

  17. 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...

  18. 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

  19. 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

  20. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-01-27

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

  2. 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.

  3. 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

  4. 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.

  5. 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...... by varying the building envelope characteristics, the power supply system, the climate, the lightning and appliances profiles, the roof tilt, the battery size and the electricity tariffs, leading to 3200 cases. The analysis is performed on an annual basis in terms of self-consumption rate, shifted energy...... 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 low feed...

  6. 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

  7. 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

  8. 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.

  9. 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.

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

    DEFF Research Database (Denmark)

    Schultz, Jørgen Munthe; Furbo, Simon

    as the theoretical results obtained for a solar combi system with the PCM-storage installed in a low energy house in a Danish climate. Parametric studies of collector area, storage volume and solar fraction for the PCM-system will be presented as well as an outline for a system with 100% coverage of the space......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...

  11. 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.

  12. 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

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  15. 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. ...

  16. 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.

  17. 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

  18. 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

  19. Radon measurements with charcoal canisters temperature and humidity considerations

    Directory of Open Access Journals (Sweden)

    Živanović Miloš Z.

    2016-01-01

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

  20. 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

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  3. 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

  4. 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)

  5. Design analysis report for the canister

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-15

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

  6. 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

  7. 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

  8. Technical note 4. Corrosion of copper canister

    International Nuclear Information System (INIS)

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

  9. 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.

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

  11. 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...

  12. 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.

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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.

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

    DEFF Research Database (Denmark)

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

    2002-01-01

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

  20. 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

  1. 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

  2. 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.

  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. 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

  5. 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

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

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

  8. 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.)

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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

  14. 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.

  15. 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.)

  16. 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

  17. 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

  18. Design of spent-fuel concrete pit dry storage and handling system

    Energy Technology Data Exchange (ETDEWEB)

    Tamaki, H.; Natsume, T.; Maruoka, K.; Yokoyama, T. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan)

    1998-07-01

    An advanced dry storage system design with highly improved storage efficiency of spent nuclear fuel has been developed. The new concept 'Concrete Pit Dry Storage System' realizes a safe and economical solution to an increasing demand of storing spent fuel assemblies (SFAs) generated from commercial nuclear power reactors. The system is basically composed of a large mass concrete module which has densely arranged pit boreholes, sealed canisters containing spent fuel assemblies and a canister handling system. The system is characterized by the following advantages compared with the existing concrete module type storage systems: higher storage efficiency can be achieved by the storage module filled with concrete which also gives a high shielding performance; simple handling technology is used for transfer and installation of the canisters at the storage facility as well as the transport cask of the canisters, surface contamination of the canister is prevented; lower radiation around the storage area is provided to reduce radiation exposure during handling and storage; high structural integrity of the facility is maintained by the concrete module with a simple construction ; the ventilation gallery introducing cooling air air to the bit borehole has an enough draft height to improve cooling performance of the system; a result of the design concept, the storage system can store higher burn-up SFAs with a short cooling period. (authors)

  19. 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

  20. 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)

  1. 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

  2. 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.

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1996-01-01

    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.

  5. Retrievable storage concept designs. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E.

    1979-03-01

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

  6. Building a Parallel Cloud Storage System using OpenStack’s Swift Object Store and Transformative Parallel I/O

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Andrew J. [Los Alamos National Laboratory; Lora, Kaleb D. [Los Alamos National Laboratory; Martinez, Esteban [Los Alamos National Laboratory; Shorter, Martel L. [Los Alamos National Laboratory

    2012-07-30

    Our project consists of bleeding-edge research into replacing the traditional storage archives with a parallel, cloud-based storage solution. It used OpenStack's Swift Object Store cloud software. It's Benchmarked Swift for write speed and scalability. Our project is unique because Swift is typically used for reads and we are mostly concerned with write speeds. Cloud Storage is a viable archive solution because: (1) Container management for larger parallel archives might ease the migration workload; (2) Many tools that are written for cloud storage could be utilized for local archive; and (3) Current large cloud storage practices in industry could be utilized to manage a scalable archive solution.

  7. 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

  8. 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.

  9. 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.

  10. 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.)

  11. 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.)

  12. 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).

  13. 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.

  14. 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.

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

    Science.gov (United States)

    Casella, Andrew M.

    2007-09-01

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

  16. 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

  17. 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

  18. 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.)

  19. 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

  20. 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

  1. 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

  2. 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

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

    Institute of Scientific and Technical Information of China (English)

    汪东旺; 李泽; 赵鑫华; 尹维东; 李志坚; 元以栋

    2012-01-01

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

  4. 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).

  5. 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).

  6. Form-Stable Phase Change Materials Based on Eutectic Mixture of Tetradecanol and Fatty Acids for Building Energy Storage: Preparation and Performance Analysis

    Directory of Open Access Journals (Sweden)

    Yiran li

    2013-10-01

    Full Text Available This paper is focused on preparation and performance analysis of a series of form-stable phase change materials (FSPCMs, based on eutectic mixtures as phase change materials (PCMs for thermal energy storage and high-density polyethylene (HDPE-ethylene-vinyl acetate (EVA polymer as supporting materials. The PCMs were eutectic mixtures of tetradecanol (TD–capric acid (CA, TD–lauric acid (LA, and TD–myristic acid (MA, which were rarely explored before. Thermal properties of eutectic mixtures and FSPCMs were measured by differential scanning calorimeter (DSC. The onset melting/solidification temperatures of form-stable PCMs were 19.13 °C/13.32 °C (FS TD–CA PCM, 24.53 °C/24.92 °C (FS TD–LA PCM, and 33.15 °C/30.72 °C (FS TD–MA PCM, respectively, and latent heats were almost greater than 90 J/g. The surface morphologies and chemical stability of form-stable PCM were surveyed by scanning electron microscopy (SEM and Fourier-transform infrared (FT-IR spectroscopy, respectively. The thermal cycling test revealed that the thermal reliability of these three form-stable PCMs was good. Thermal storage/release experiment indicated melting/solidification time was shortened by introducing 10 wt % aluminum powder (AP. It is concluded that these FSPCMs can act as potential building thermal storage materials in terms of their satisfactory thermal properties.

  7. The Unity connecting module is moved to payload canister

    Science.gov (United States)

    1998-01-01

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

  8. 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.

  9. 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)

  10. 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

  11. 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.

  12. Reliability in sealing of canister for spent nuclear fuel

    International Nuclear Information System (INIS)

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

  13. Reliability in sealing of canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

  14. Effects of brine migration on waste storage systems. Final report. [Thermomechanical effects

    Energy Technology Data Exchange (ETDEWEB)

    Gaffney, E.S.; Nickell, R.E.

    1979-05-15

    Processes which can lead to mobilization of brine adjacent to spent fuel or nuclear waste canisters and some of the thermomechanical consequences have been investigated. Velocities as high as 4 x 10/sup -7/ m s/sup -1/ (13 m y/sup -1/) are calculated at the salt/canister boundary. As much as 40 liters of pure NaCl brine could accumulate around each canister during a 10-year storage period. Accumulations of bittern brines would probably be less, in the range of 2 to 5 liters. With 0.5% water, NaCl brine accumulation over a 10-year storage cycle around a spent fuel canister producing 0.6 kW of heat is expected to be less than 1 liter for centimeter-size inclusions and less than 0.5 liter for millimeter-size inclusions. For bittern brines, about 25 years would be required to accumulate 0.4 liter. The most serious mechanical consequence of brine migration would be the increased mobility of the waste canister due to pressure solution. In pressure solution enhanced deformation, the existence of a thin film of fluid either between grains or between media (such as between a canister and the salt) provides a pathway by which the salt can be redistributed leading to a marked increase in strain rates in wet rock relative to dry rock. In salt, intergranular water will probably form discontinuous layers rather than films so that they would dominate pressure solution. A mathematical model of pressure solution indicates that pressure solution will not lead to appreciable canister motions except possibly in fine grained rocks (less than 10/sup -4/ m). In fine grained salts, details of the contact surface between the canister and the salt bed may lead to large pressure solution motions. A numerical model indicates that heat transfer in the brine layer surrounding a spent fuel canister is not conduction dominated but has a significant convective component.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SHERRELL, D.L.

    2000-10-10

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

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

    International Nuclear Information System (INIS)

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

  19. Procedural development for nuclear waste canister impact testing

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2005-11-01

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

  1. Preselection of Ni-Cr(-Mo) alloys as potential canister materials for vitrified high active nuclear waste by electrochemical testing

    Science.gov (United States)

    Bort, H.; Wolf, I.; Leistikow, S.

    1987-07-01

    Several Ni-Cr(-Mo) alloys (Hastelloy C4, Inconel 625, Sanicro 28, Incoloy 825, Inconel 690) were tested by electrochemical methods to characterize their corrosion behavior in chloride containing solutions at various temperatures and pH-values in respect to their application as canister materials for final radioactive waste storage. Especially, Hastelloy C4 was tested by potentiodynamic, potentiostatic and galvanostic measurements. As electrolytes H 2SO 4 solutions were used, as parameters temperature, chloride content and pH-value were varied. All tested alloys showed a clearly limited resistance against pitting corrosion phenomena; under severe conditions even crevice corrosion phenomena were observed. The best corrosion behavior, however, is shown by Hastelloy C4, which has the lowest passivation current density of all tested alloys and the largest potential region with protection against local corrosion phenomena.

  2. Preselection of Ni-Cr(-Mo) alloys as potential canister materials for vitrified high active nuclear waste by electrochemical testing

    Energy Technology Data Exchange (ETDEWEB)

    Bort, H.; Wolf, I.; Leistikow, S.

    1987-07-01

    Several Ni-Cr(-Mo) alloys (Hastelloy C4, Inconel 625, Sanicro 28, Incoloy 825, Inconel 690) were tested by electrochemical methods to characterize their corrosion behavior in chloride containing solutions at various temperatures and pH-values in respect to their application as canister materials for final radioactive waste storage. Especially, Hastelloy C4 was tested by potentiodynamic, potentiostatic and galvanostic measurements. As electrolytes H/sub 2/SO/sub 4/ solutions were used, as parameters temperature, chloride content and pH-value were varied. All tested alloys showed a clearly limited resistance against pitting corrosion phenomena; under severe conditions even crevice corrosion phenomena were observed. The best corrosion behavior, however, is shown by Hastelloy C4, which has the lowest passivation current density of all tested alloys and the largest potential region with protection against local corrosion phenomena.

  3. 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)

  4. Choices of canisters and elements for the first fuel shipment from K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-03-01

    Twenty-two canisters (10 prime and 12 backup candidates) in the K West Basin have been identified as containing fuel which, when examined, will satisfy the Data Quality Objectives for the first fuel shipment from this basin. These were chosen as meeting criteria such as containing relatively long fuel elements, locking bar integrity, and the availability of gas/liquid interface level measurements for associated canister gas traps. Two canisters were identified as having reported broken fuel on initial loading. Usage and interpretation of canister cesium concentration measurements have also been established and levels of maximum and minimum acceptable cesium concentration (from a data optimization point of view) for decapping have been determined although other operational cesium limits may also apply. Criteria for picking particular elements, once a canister is opened, are reviewed in this document. A pristine, a slightly damaged, and a badly damaged element are desired. The latter includes elements with end caps removed but does not include elements which have large amounts of swelling or split cladding that might interfere with handling tools. Finally, operational scenarios have been suggested to aid in the selections of canisters and elements in a way that utilizes anticipated canister gas sampling and leads to a correct and quick choice of elements which will supply the desired data.

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hutsell, D.J.

    2000-02-29

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Cannell, G.R.

    1999-10-07

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

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

    International Nuclear Information System (INIS)

    Copper canister corrosion is an important issue in the concept of a nuclear fuel repository. Previous studies indicate that the oxygen-free copper canister could hold its integrity for more than 100 000 years in the repository environment. Microbial processes may reduce sulphate to sulphide and considerably increase the amount of sulphide available for corrosion. In this paper, a coupled transport/reaction model is developed to account for the transport of chemical species produced by microbial processes. The corroding agents like sulphide would come not only from the intruding groundwater, but also from the reduction of sulphate near the canister. The reaction of sulphate-reducing bacteria and the transport of sulphide in the bentonite buffer is included in the model. The local depth of copper canister corrosion is calculated by the model. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

    Copper canister corrosion is an important issue in the concept of a nuclear fuel repository. Previous studies indicate that the oxygen-free copper canister could hold its integrity for more than 100 000 years in the repository environment. Microbial processes may reduce sulphate to sulphide and considerably increase the amount of sulphide available for corrosion. In this paper, a coupled transport/reaction model is developed to account for the transport of chemical species produced by microbial processes. The corroding agents like sulphide would come not only from the intruding groundwater, but also from the reduction of sulphate near the canister. The reaction of sulphate-reducing bacteria and the transport of sulphide in the bentonite buffer is included in the model. The local depth of copper canister corrosion is calculated by the model. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-01

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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-03-01

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

  19. Certification of a spent fuel cask for storage and transportation

    International Nuclear Information System (INIS)

    This paper addresses the US Nuclear Regulatory Commission's requirements for the dry storage and transportation of spent fuel, focusing on how the performance standards differ between storage and transportation. The paper also discusses the NRC cask review process, and some current issues in each area of certification. In addition, some of the issues associated with the US Department of Energy's proposed multi-purpose canister are discussed

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

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

  1. Design basis for the copper canister. Stage one

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-02-01

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

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

    Science.gov (United States)

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

    2007-05-01

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

  3. FACSIM/MRS [Monitored Retrievable Storage]-2: Storage and shipping model documentation and user's guide

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory (PNL) has developed a stochastic computer model, FACSIM/MRS, to assist in assessing the operational performance of the Monitored Retrievable Storage (MRS) waste-handling facility. This report provides the documentation and user's guide for FACSIM/MRS-2, which is also referred to as the back-end model. The FACSIM/MRS-2 model simulates the MRS storage and shipping operations, which include handling canistered spent fuel and secondary waste in the shielded canyon cells, in onsite yard storage, and in repository shipping cask loading areas

  4. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, R.G.

    1983-08-01

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

  6. Dry storage of the BR3 spent fuel in the Castor BR3 cask

    Energy Technology Data Exchange (ETDEWEB)

    Ooms, L.; Massaut, V.; Noynaert, L. [SCK/CEN, B-2400 Mol (Belgium); Braeckeveldt, M. [Niras/Ondraf, B-1210 Brussels (Belgium)

    2001-07-01

    Twenty-five years of operation has resulted in an inventory of spent fuel with a wide variety in the BR3 nuclear pilot power plant. Studies were launched to evaluate all possible solutions for the BR3 experimental and 'exotic' spent fuel, i.e. reprocessing, dry storage in containers and dry storage in canisters. For the BR3 spent fuel the interim dry storage in Castor BR3 containers was chosen. The present paper describes in a first part the history and characteristics of the spent fuel. A second part handles with the different options, which were studied for the spent fuel evacuation. The last part focuses on the spent fuel preparation and the production of the Castor BR3. This project allowed the SCK-CEN to build up an important know-how in the field of spent fuel management, especially the management of research reactor fuel, which is very specific and not comparable with spent fuel of commercial nuclear power plants. (author)

  7. Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

    International Nuclear Information System (INIS)

    Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in

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

    Science.gov (United States)

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

    2015-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-06-01

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

  10. Manufacturing of the canister shells T54 and T55

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. Effects of long term dry storage of spent fuel on the performance of further extended storage, transport and disposal packaging

    International Nuclear Information System (INIS)

    In this paper the performance of spent fuel is analyzed on the basis of 3 categories of storage modes, which depend on the temperature cooling time relationship of the fuel. Sufficient data are available to prove that storage periods of at least 50 years are safe. For extended storage periods the mechanical strength of the fuel is sufficient for all normal handling operations, whereas fuel rod integrity plays a minor role. Extended storage in mode II and III for 50-150 years needs additional research in selected areas to reduce uncertainties in extrapolating over long periods of time. Alternatively uncertainties can be covered by using robust canister concepts

  13. Spent unreprocessed fuel (SURF) facility evaluation plan of the alternative storage concepts

    International Nuclear Information System (INIS)

    Concepts were evaluated for the storage of unreprocessed spent fuel in a retrievable surface storage facility. This document provides a systematic format for making a concept selection from the seven alternative concepts presented in RHO-LD-2. Results of the evaluation was that the Drywell concept was rated highest with the Water Basin Concept and the Sealed Storage Cask concept with multiple canisters of SURF coming in a close second and third

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

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-03-09

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-08-01

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

  20. Radiation Heat Transfer Modeling Improved for Phase-Change, Thermal Energy Storage Systems

    Science.gov (United States)

    Kerslake, Thomas W.; Jacqmin, David A.

    1998-01-01

    Spacecraft solar dynamic power systems typically use high-temperature phase-change materials to efficiently store thermal energy for heat engine operation in orbital eclipse periods. Lithium fluoride salts are particularly well suited for this application because of their high heat of fusion, long-term stability, and appropriate melting point. Considerable attention has been focused on the development of thermal energy storage (TES) canisters that employ either pure lithium fluoride (LiF), with a melting point of 1121 K, or eutectic composition lithium-fluoride/calcium-difluoride (LiF-20CaF2), with a 1040 K melting point, as the phase-change material. Primary goals of TES canister development include maximizing the phase-change material melt fraction, minimizing the canister mass per unit of energy storage, and maximizing the phase-change material thermal charge/discharge rates within the limits posed by the container structure.

  1. Corrosion of the copper canister in the repository environment

    International Nuclear Information System (INIS)

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

  2. Corrosion of the copper canister in the repository environment

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Soo; Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won [Korea Atomic Energy Institute, Daejeon (Korea, Republic of)

    2012-09-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-15

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

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

    International Nuclear Information System (INIS)

    Copper canister corrosion is an important issue in the concept of a nuclear fuel repository. Previous studies indicate that the oxygen-free copper canister could hold its integrity for more than 100,000 years in the repository environment. Microbial processes may reduce sulphate to sulphide and considerably increase the amount of sulphides available for corrosion. In this paper, a coupled transport/reaction model is developed to account for the transport of chemical species produced by microbial processes. The corroding agents like sulphide would come not only from the groundwater flowing in a fracture that intersects the canister, but also from the reduction of sulphate near the canister. The reaction of sulphate-reducing bacteria and the transport of sulphide in the bentonite buffer are included in the model. The depth of copper canister corrosion is calculated by the model. With representative 'central values' of the concentrations of sulphate and methane at repository depth at different sites in Fennoscandian Shield the corrosion depth predicted by the model is a few millimetres during 105 years. As the concentrations of sulphate and methane are extremely site-specific and future climate changes may significantly influence the groundwater compositions at potential repository sites, sensitivity analyses have been conducted. With a broad perspective of the measured concentrations at different sites in Sweden and in Finland, and some possible mechanisms (like the glacial meltwater intrusion and interglacial seawater intrusion) that may introduce more sulphate into the groundwater at intermediate depths during future climate changes, higher concentrations of either/both sulphate and methane than what is used as the representative 'central' values would be possible. In worst cases. locally, half of the canister thickness could possibly be corroded within 105 years

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-01

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

  13. STS-105 ICC is moved to the payload canister for transport to pad 39A

    Science.gov (United States)

    2001-01-01

    KENNEDY SPACE CENTER, Fla. -- An overhead crane in the Space Station Processing Facility lifts the Integrated Cargo Carrier from its workstand to move it to the payload canister. The ICC holds several payloads for mission STS-105, the Early Ammonia Servicer and two experiment containers. The ICC will join the Multi-Purpose Logistics Module Leonardo in the payload canister for transport to Launch Pad 39A where they will be placed in the payload bay of Space Shuttle Discovery. Launch of STS-105 is scheduled for 5:38 p.m. EDT Aug. 9

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Herrington, Jason S

    2013-08-20

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

  17. Surface storage cask test summarization report

    International Nuclear Information System (INIS)

    From December 1978 to September 1982, as part of DOE's Spent Fuel Handling and Packaging Program and Commercial Waste and Spent Fuel Packaging Program, a pressurized water reactor (PWR) spent nuclear fuel assembly with an initial decay heat level of approximately 1.0 kilowatt (kW) was emplaced in a concrete cask at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility in Area 25 of the Nevada Test Site. Temperatures were monitored during the emplacement period to determine the thermal response of the cask, the canister, and the fuel assembly. During and following the test, the atmosphere of the canister containing the fuel assembly was sampled to determine if fission product gases had been released by the fuel assembly. This 45-month Surface Storage Cask (SSC) test was the first demonstration of interim storage of a PWR spent fuel assembly in a dry storage cask. The receipt, handling, packaging, emplacement and retrieval operations have been demonstrated as directly applicable to similar operations in federal interim storage and repository related activities. 7 references, 35 figures, 7 tables

  18. Energy Storage.

    Science.gov (United States)

    Eaton, William W.

    Described are technological considerations affecting storage of energy, particularly electrical energy. The background and present status of energy storage by batteries, water storage, compressed air storage, flywheels, magnetic storage, hydrogen storage, and thermal storage are discussed followed by a review of development trends. Included are…

  19. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    Energy Technology Data Exchange (ETDEWEB)

    HUMPHREYS, D C

    2002-08-01

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  1. Spent Nuclear Fuel (SNF) Project Multi Canister Overpack (MCO) Process Flow Diagram Mass Balance Calculations

    Energy Technology Data Exchange (ETDEWEB)

    KLEM, M.J.

    2000-09-08

    The purpose of this calculation document is to develop the bases for the material balances of the Multi-Canister Overpack (MCO) Level 1 Process Flow Diagram (PFD). The attached mass balances support revision two of the PFD for the MCO and provide future reference.

  2. Corrosion issues in relation to copper canisters for disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Saario, T. (VTT Technical Research Centre of Finland, Espoo (Finland))

    2010-05-15

    Spent nuclear fuel in Finland and Sweden will be disposed in deposition holes excavated in granitic bedrock at a depth of about 400 to 500 m. The release of activated products is prevented by a multi-barrier concept. A copper canister with 50 mm wall thickness will be the main corrosion shield, expected to last un-perforated for over 100 000 years. There are several different corrosion mechanisms that in principle may threaten the integrity of the copper canister. Likewise, there are several different scenarios which may expose the copper canister to these corrosion phenomena. During the years, most of the corrosion related concerns have been satisfactorily dealt with, leaving a few to be further researched. This paper gives an overview of the corrosion related issues with regard to the integrity of the copper canister, as well as the main approaches developed by the share holders to ensure the integrity for the duration needed. A more detailed description of the few remaining ongoing corrosion related research items is also included. (orig.)

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

    International Nuclear Information System (INIS)

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

  4. 40 CFR 86.153-98 - Vehicle and canister preconditioning; refueling test.

    Science.gov (United States)

    2010-07-01

    ... controlled to 50±25 grains of water vapor per pound of dry air) maintained at a nominal flow rate of 0.8 cfm... at least 1200 canister bed volumes of ambient air (with humidity controlled to 50±25 grains of water...'s air conditioner (if so equipped) shall be turned off. Ambient temperature shall be controlled...

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

    International Nuclear Information System (INIS)

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

  6. Reliability in NDT of canister for the Swedish spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ronneteg, Ulf [SKB, Oskarshamn (Sweden); Mueller, Christina; Pavlovic, Mato [BAM, Berlin (Germany)

    2009-07-01

    The Swedish KBS-3 design for the disposal of spent fuel is based on encapsulation of the fuel in canisters consisting of cast iron inserts and an outer 5 cm thick shield of copper. The canisters are embedded in bentonite clay and will be disposed in crystalline bedrock at a depth of about 500 m. To verify that the canisters fulfil the requirements, an extensive program for quality control is developed. In this program the use of NDT is vital and therefore it is very important to determine the reliability of the developed NDT methods. The reliability of the NDT methods has been studied for six years and the results are used in several ways. One task is to identify areas where further optimizations of the NDT methods are necessary and another is to study the effect of influencing parameters prior to the upcoming qualifications. Furthermore, the reliability analyses of the NDT are combined with analyses of the manufacturing and welding processes to give input to the overall safety analyses of the long-term properties of the canisters. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars; Lilja, Christina (eds.)

    2010-12-15

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

  10. Environmental and Economic Assessment of Reclaimed Polyurethane Panels: The Case of Diverting Decommissioned Cold Storage Panels From Landfills and Recycling Into Three Forms of Insulative Building Materials

    OpenAIRE

    Costanza, James

    2015-01-01

    This study investigates the long-term thermal performance of polyurethane insulated cold storage panels and the environmental and economic impact of recycling such panels when taken out of service in lieu of discarding them in landfills. It is estimated, as of 2015, over 180 million square feet of insulated cold storage panels are manufactured annually in the U.S. The panels are most frequently constructed of closed-cell, low density polyurethane insulation utilizing HCF 245fa and HCF 1...

  11. Macro-encapsulation of heat storage phase-change materials for use in residential buildings. First quarterly progress report, September 29--December 29, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Lane, G.A.; Kott, A.C.; Rossow, H.E.

    1977-02-01

    Objectives are to assess the feasibility of macro-encapsulated PCM's for residential solar systems, to develop and evaluate such materials. Five PCM's have been selected from encapsulation studies. Encapsulated storage media were evaluated theoretically in storage beds with air and water as the heat transfer medium. Cylindrical, tetrahedral, and pillow shapes are being evaluated for the encapsulated PCM. Encapsulant materials under consideration are multilayer flexible plastic films, steel cans, and plastic bottles.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Multipurpose dry storage facilities

    International Nuclear Information System (INIS)

    SGN has gained considerable experience in the design and construction of interim storage facilities for spent fuel and various nuclear waste, and can therefore propose single product and multiproduct facilities capable of accommodating all types of waste in a single structure. The pooling of certain functions (transport cask reception, radiation protection), the choice of optimized technologies to meet the specific needs of the clients (automatic transfer by shielded cask or nuclearized crane) and the use of the same type of well to cool the heat releasing packages (glass canisters, fuel elements) make it possible to propose industrially proven and cost effective solutions. The studies conducted by SGN on behalf of the Dutch company COVRA (Centrale Organisatie Voor Radioactif Afval), offer an example of the application of this new concept. This paper first presents the SGN experience through a short description of reference storage facilities for various types of products (MLW, HLW and Spent Fuel). It goes on with a typical application to show how these proven technologies are combined to obtain single product or multiproduct facilities tailored to the client's specific requirements. (author)

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

    Science.gov (United States)

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

    2007-01-01

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

  19. Spent Fuel Dry Storage Cask Thermal Test

    International Nuclear Information System (INIS)

    Most nuclear power plants maintain their spent fuel discharged at a reactor in wet storage pools. However, after several years of use, many pools are filled to capacity. Therefore, finding a sufficient capacity for storage is essential because of the continued delays in obtaining a safe, interim storage facility if nuclear power plants are to be allowed to continue to operate. Dry storage cask will be one solution for solving an interim storage problem. The dry storage cask consists of two separate components: an over-pack, and a canister. The structure strength part of the over-pack is made of carbon steel, and the inner cavity of the structure strength part is filled with concrete, which accomplishes the role as a radiation shield. The outer diameter of the dry storage cask is 3,550 mm and the its overall height is 5,885 mm. It weighs approximately 135 tons. The dry storage cask accommodates 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat from the 24 PWR spent fuel assemblies is 25.2 kW This paper discusses the experimental approach used to evaluate the heat transfer characteristics of the dry storage cask

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-03-01

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

  1. Low Oxygen Storage of Farmer Stock Peanuts

    Science.gov (United States)

    Farmer stock peanuts are stored in bulk storage facilities for periods ranging from 30d to 12mo. Studies were conducted in 1/10 scale conventional and monolithic dome storage facilities located in Dawson, GA. Conventional storage was represented by four metal buildings with storage capacity of appro...

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

    International Nuclear Information System (INIS)

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

  3. CFD Simulation of Spent Fuel in a Dry Storage System

    International Nuclear Information System (INIS)

    The spent fuel pool is expected to be full in few years. It is a serious problem one should not ignore. The dry storage type is considered as the interim storage system in Korea. The system stores spent fuel in a storage canister filled with an inert gas and the canister is cooled by a natural convection system using air or helium, radiation, and conduction. The spent fuel is heated by decay heat. The spent fuel is allowed to cool under a limiting temperature to avoid a fuel failure. Recently, the thermal hydraulic characteristics for a single bundle of the spent fuel were investigated through a CFD simulation. It would be of great interest to investigate the maximum fuel temperature in a dry storage system. The present paper deals with the thermal hydraulic characteristics of spent fuel for a dry storage system using the CFD method. A 3-D thermal flow simulation was carried out to predict the temperature of spent fuel. A dry storage system composed of 32 fuel bundles was modeled. The inlet temperature of the outer bundle is higher and that of inner bundle, however, is higher at the outlet. In a single fuel assembly, a center temperature of the fuel assembly was higher than elsewhere

  4. CFD Simulation of Spent Fuel in a Dry Storage System

    Energy Technology Data Exchange (ETDEWEB)

    Kwack, Young Kyun; In, Wang Kee; Shin, Chang Hwan; Chun, Tae Hyun; Kook, Dong Hak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The spent fuel pool is expected to be full in few years. It is a serious problem one should not ignore. The dry storage type is considered as the interim storage system in Korea. The system stores spent fuel in a storage canister filled with an inert gas and the canister is cooled by a natural convection system using air or helium, radiation, and conduction. The spent fuel is heated by decay heat. The spent fuel is allowed to cool under a limiting temperature to avoid a fuel failure. Recently, the thermal hydraulic characteristics for a single bundle of the spent fuel were investigated through a CFD simulation. It would be of great interest to investigate the maximum fuel temperature in a dry storage system. The present paper deals with the thermal hydraulic characteristics of spent fuel for a dry storage system using the CFD method. A 3-D thermal flow simulation was carried out to predict the temperature of spent fuel. A dry storage system composed of 32 fuel bundles was modeled. The inlet temperature of the outer bundle is higher and that of inner bundle, however, is higher at the outlet. In a single fuel assembly, a center temperature of the fuel assembly was higher than elsewhere.

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

    OpenAIRE

    Liu, Longcheng

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

  7. Localization of plastic deformation in copper canisters for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, K.; Saukkonen, T.; Haenninen, H. (Aalto Univ. School of Science and Technology, Espoo (Finland))

    2010-05-15

    In Finland and Sweden the spent nuclear fuel will be deposited in a deep repository in copper corrosion barrier canisters surrounding cast iron inserts. The 50 mm thick copper canisters will we sealed using either electron beam welding (EB) or friction stir welding (FSW) to join the tubes and the lids/bottoms. The canisters will deform in the repository conditions, e.g., due to the hydrostatic pressure. The deformation will localize to different discontinuities, such as defects and microstructural heterogeneity. This study compared the localization of plastic deformation in EB and FSW welds as well as in the base materials (both forged and extruded) using optical strain measurement methods. The results show that in the base materials the deformation occurs very uniformly over the entire gauge length. In FSW welds the deformation localizes in the weld at a line of entrapped oxide particles. In EB welds the deformation localizes to the large grains in the middle of the weld or at the steep grain size gradient between the weld and the tube material. Tensile strength is lowest in the EB welds (175 MPa as compared to 200 MPa or higher for the other samples). Elongation to fracture of the FSW welds is similar to those of the base materials, but for the EB welds it is significantly lower (appr. 40 % compared to 65 %). (orig.)

  8. Inner material requirements and candidates screening for spent fuel disposal canister

    International Nuclear Information System (INIS)

    In the context of the present Spanish 'once-through' nuclear fuel cycle, the need arises to complete the geological repository reference concept with a spent fuel canister final design. One of the main issues in its design is selecting the inner material to be placed inside the canister, between the steel walls and the spent fuel assemblies. The primary purpose of this material will be to avoid the possibility of a criticality event once the canister walls have been finally breached by corrosion and the spent fuel is flooded with groundwater. That is an important role because the increase in heat generation from such an event would act against spent fuel stability and compromise bentonite barrier functions, negatively affecting overall repository performance. To prevent this possibility a detailed set of requirements for a material to fulfil this role in the repository environment have been devised and presented in this paper. With these requirements in view, eight potentially interesting candidates were selected and evaluated: cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, haematite, phosphates, and olivine. Among these, the first four materials or material families are found promising for this application. In addition, other relevant non-performance-related aspects of candidate materials, which could help on decision making, are also considered and evaluated. (authors)

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

    Science.gov (United States)

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

    2006-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chad Pope; Larry L. Taylor; Soon Sam Kim

    2007-02-01

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

  11. New Strategies for Licensing the Storage and Transportation of High Burn-up Spent Nuclear Fuel in the United States - 12546

    International Nuclear Information System (INIS)

    An alternative approach may be needed to the licensing of high-burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates. Packaging strategies and regulations should be developed to reduce the potential for requiring fuel to be repackaged unnecessarily. This would lessen the chance of accidents and mishaps during loading and unloading of casks, and decrease dose to workers. A packaging approach that shifts the safety basis from reliance upon the fuel condition to reliance upon an inner canister could eliminate or lessen the need for repackaging. In addition, the condition of canisters can be more readily monitored and inspected than the condition of fuel cladding. Canisters can also be repaired and/or replaced when deemed necessary. In contrast, once a fuel assembly is loaded into a canister and placed in a storage overpack, there is little opportunity to monitor its condition or take mitigating measures if cladding degradation is suspected or proven to occur. (authors)

  12. Performance Assessment and Sensitivity Analyses of Disposal of Plutonium as Can-in-Canister Ceramic

    Energy Technology Data Exchange (ETDEWEB)

    Rainer Senger

    2001-09-25

    The purpose of this analysis is to examine whether there is a justification for using high-level waste (HLW) as a surrogate for plutonium disposal in can-in-canister ceramic in the total-system performance assessment (TSPA) model for the Site Recommendation (SR). In the TSPA-SR model, the immobilized plutonium waste form is not explicitly represented, but is implicitly represented as an equal number of canisters of HLW. There are about 50 metric tons of plutonium in the U. S. Department of Energy inventory of surplus fissile material that could be disposed. Approximately 17 tons of this material contain significant quantities of impurities and are considered unsuitable for mixed-oxide (MOX) reactor fuel. This material has been designated for direct disposal by immobilization in a ceramic waste form and encapsulating this waste form in high-level waste (HLW). The remaining plutonium is suitable for incorporation into MOX fuel assemblies for commercial reactors (Shaw 1999, Section 2). In this analysis, two cases of immobilized plutonium disposal are analyzed, the 17-ton case and the 13-ton case (Shaw et al. 2001, Section 2.2). The MOX spent-fuel disposal is not analyzed in this report. In the TSPA-VA (CRWMS M&O 1998a, Appendix B, Section B-4), the calculated dose release from immobilized plutonium waste form (can-in-canister ceramic) did not exceed that from an equivalent amount of HLW glass. This indicates that the HLW could be used as a surrogate for the plutonium can-in-canister ceramic. Representation of can-in-canister ceramic as a surrogate is necessary to reduce the number of waste forms in the TSPA model. This reduction reduces the complexity and running time of the TSPA model and makes the analyses tractable. This document was developed under a Technical Work Plan (CRWMS M&O 2000a), and is compliant with that plan. The application of the Quality Assurance (QA) program to the development of that plan (CRWMS M&O 2000a) and of this Analysis is described in

  13. Monitored Retrievable Storage System Requirements Document

    International Nuclear Information System (INIS)

    This Monitored Retrievable Storage System Requirements Document (MRS-SRD) describes the functions to be performed and technical requirements for a Monitored Retrievable Storage (MRS) facility subelement and the On-Site Transfer and Storage (OSTS) subelement. The MRS facility subelement provides for temporary storage, at a Civilian Radioactive Waste Management System (CRWMS) operated site, of spent nuclear fuel (SNF) contained in an NRC-approved Multi-Purpose Canister (MPC) storage mode, or other NRC-approved storage modes. The OSTS subelement provides for transfer and storage, at Purchaser sites, of spent nuclear fuel (SNF) contained in MPCs. Both the MRS facility subelement and the OSTS subelement are in support of the CRWMS. The purpose of the MRS-SRD is to define the top-level requirements for the development of the MRS facility and the OSTS. These requirements include design, operation, and decommissioning requirements to the extent they impact on the physical development of the MRS facility and the OSTS. The document also presents an overall description of the MRS facility and the OSTS, their functions (derived by extending the functional analysis documented by the Physical System Requirements (PSR) Store Waste Document), their segments, and the requirements allocated to the segments. In addition, the top-level interface requirements of the MRS facility and the OSTS are included. As such, the MRS-SRD provides the technical baseline for the MRS Safety Analysis Report (SAR) design and the OSTS Safety Analysis Report design

  14. Thermal Energy Storage: Fourth Annual Review Meeting

    Science.gov (United States)

    1980-01-01

    The development of low cost thermal energy storage technologies is discussed in terms of near term oil savings, solar energy applications, and dispersed energy systems for energy conservation policies. Program definition and assessment and research and technology development are considered along with industrial storage, solar thermal power storage, building heating and cooling, and seasonal thermal storage. A bibliography on seasonal thermal energy storage emphasizing aquifer thermal energy is included.

  15. UFD Storage and Transportation - Transportation Working Group Report

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Ross, Steven B.

    2011-08-01

    The Used Fuel Disposition (UFD) Transportation Task commenced in October 2010. As its first task, Pacific Northwest National Laboratory (PNNL) compiled a list of structures, systems, and components (SSCs) of transportation systems and their possible degradation mechanisms during extended storage. The list of SSCs and the associated degradation mechanisms [known as features, events, and processes (FEPs)] were based on the list of used nuclear fuel (UNF) storage system SSCs and degradation mechanisms developed by the UFD Storage Task (Hanson et al. 2011). Other sources of information surveyed to develop the list of SSCs and their degradation mechanisms included references such as Evaluation of the Technical Basis for Extended Dry Storage and Transportation of Used Nuclear Fuel (NWTRB 2010), Transportation, Aging and Disposal Canister System Performance Specification, Revision 1 (OCRWM 2008), Data Needs for Long-Term Storage of LWR Fuel (EPRI 1998), Technical Bases for Extended Dry Storage of Spent Nuclear Fuel (EPRI 2002), Used Fuel and High-Level Radioactive Waste Extended Storage Collaboration Program (EPRI 2010a), Industry Spent Fuel Storage Handbook (EPRI 2010b), and Transportation of Commercial Spent Nuclear Fuel, Issues Resolution (EPRI 2010c). SSCs include items such as the fuel, cladding, fuel baskets, neutron poisons, metal canisters, etc. Potential degradation mechanisms (FEPs) included mechanical, thermal, radiation and chemical stressors, such as fuel fragmentation, embrittlement of cladding by hydrogen, oxidation of cladding, metal fatigue, corrosion, etc. These degradation mechanisms are discussed in Section 2 of this report. The degradation mechanisms have been evaluated to determine if they would be influenced by extended storage or high burnup, the need for additional data, and their importance to transportation. These categories were used to identify the most significant transportation degradation mechanisms. As expected, for the most part, the

  16. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    In this report the various facilities incorporated in the proposed handling chain for spent fuel from the power stations to the final repository are discribed. Thus the geological conditions which are essential for a final repository is discussed as well as the buffer and canister materials and how they contribute towards a long-term isolation of the spent fuel. Furthermore one chapter deals with leaching of the deposited fuel in the event that the canister is penetrated as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material. The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the ecological systems as well as radiation doses. Finally a summary is given of the safety analysis of the proposed method for the handling and final storage of the spent fuel. (E.R.)

  17. Safety Tests of Concrete Storage Cask for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    In preparation for the timely installation of interim storage facility for spent nuclear fuel (SF), KORAD is developing domestic models of SF storage systems and the concrete storage cask is one of them. A concrete cask consists of a metallic canister which confines SF with welded closure and a concrete overpack which provides radiation shielding and physical protection to the canister. The safety requirements for a SF storage cask is well established in US and summarized in regulatory guides such as NUREG-1536. KAERI has been performing tests of the concrete cask to demonstrate its safety and compliance to the regulatory requirements with high priority stipulated in NUREG-1536. The test program includes the structural performance tests under tip-over and earthquake and decay heat removal test under normal, off-normal and accident conditions. In this paper, brief introduction to the structural tests and their results are provided. Safety tests to demonstrate the safety of KORAD21C concrete storage cask were successfully performed. The structural integrity during tip-over and earthquake were demonstrated with scale model tests and the results are analyzed in comparison with safety analysis results

  18. Analysis of thermal energy storage material with change-of-phase volumetric effects

    Science.gov (United States)

    Kerslake, Thomas W.; Ibrahim, Mounir B.

    1990-01-01

    NASA's Space Station Freedom proposed hybrid power system includes photovoltaic arrays with nickel hydrogen batteries for energy storage and solar dynamic collectors driving Brayton heat engines with change-of-phase Thermal Energy Storage (TES) devices. A TES device is comprised of multiple metallic, annular canisters which contain a eutectic composition LiF-CaF2 Phase Change Material (PCM) that melts at 1040 K. A moderately sophisticated LiF-CaF2 PCM computer model is being developed in three stages considering 1-D, 2-D, and 3-D canister geometries, respectively. The 1-D model results indicate that the void has a marked effect on the phase change process due to PCM displacement and dynamic void heat transfer resistance. Equally influential are the effects of different boundary conditions and liquid PCM natural convection. For the second stage, successful numerical techniques used in the 1-D phase change model are extended to a 2-D (r,z) PCM containment canister model. A prototypical PCM containment canister is analyzed and the results are discussed.

  19. Irradiated fuel storage program in Ontario Hydro

    International Nuclear Information System (INIS)

    The primary objective of Ontario Hydro's irradiated fuel storage program is to manage the used fuel discharged from their existing and future nuclear generating stations in a safe and economical manner until it is ready for reprocessing or disposal. While current irradiated fuel handling and storage systems are adequate, some modifications will be required to integrate them with eventual transportation and disposal systems. A secondary objective of the storage program, therefore, is to review and modify existing facilities and to develop future facilities using a total systems approach. Alternate storage concepts are being evaluated as part of a study which looks into the feasibility of storing irradiated fuel for periods greater than 50 years. The main requirements considered for extended storage of irradiated fuel are long-term retrievability, long-lasting structures, and long-term maintainability and control. The basic criteria governing the study are: safety, responsibility, current technology, environmental impact, economics, and public and social acceptability. The extended storage concepts considered are: (1) engineered air- or water-cooled vaults aboveground, near surface, or deep underground; (2) borehole emplacement aboveground, near surface, or deep underground; and (3) concrete canisters aboveground. Technical descriptions of the concepts from which the assessments will be made are now being prepared. So-called hybrid storage systems are also being investigated. The hybrid strategies consider changing from a wet storage system to a dry storage system, relocation from at surface to underground, and/or changing from an active cooling system to a passive cooling system after an initial storage period of 100 to 200 years. The final report, with evaluations of extended storage strategies, is scheduled for completion in early 1981

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

  1. A supermolecular building layer approach for gas separation and storage applications: The eea and rtl MOF platforms for CO2 capture and hydrocarbon separation

    KAUST Repository

    Chen, Zhijie

    2015-01-01

    The supermolecular building layer (SBL) approach was employed to deliberately synthesize five novel metal-organic frameworks (1-5) with an exposed array of amide or amine functionalities within their pore system. The ability to decorate the pores with nitrogen donor moieties offers potential to evaluate/elucidate the structure-adsorption property relationship. Two MOF platforms, eea-MOF and rtl-MOF, based on pillaring of kgm-a or sql-a layers with heterofunctional 3-connected organic building blocks were targeted and constructed to purposely introduce and expose the desired amide or amine functionalities. Interestingly, gas adsorption properties of eea-MOF-4 (1) and eea-MOF-5 (2) showed that by simply altering the nitrogen donor position within the ligand, it is possible to relatively reduce the pore size of the related eea-MOF material and subsequently increase the associated CO2 uptake. The slightly confined pore space in 2, relative to 1, has enabled an enhancement of the pore local charge density and thus the observed relative increase in the CO2 and H2 isosteric heat of adsorption (Qst). In addition, light hydrocarbon adsorption studies revealed that 2 is more selective toward C2H6 and C3H8 over CH4 than 1, as exemplified for C2H6:CH4 (5:95) or C3H8:CH4 (5:95) binary gas mixtures. This journal is

  2. A supermolecular building layer approach for gas separation and storage applications: the eea and rtl MOF platforms for CO 2 capture and hydrocarbon separation

    KAUST Repository

    Chen, Zhijie

    2015-02-11

    The supermolecular building layer (SBL) approach was employed to deliberately synthesize five novel metal–organic frameworks (1–5) with an exposed array of amide or amine functionalities within their pore system. The ability to decorate the pores with nitrogen donor moieties offers potential to evaluate/elucidate the structure–adsorption property relationship. Two MOF platforms, eea-MOF and rtl-MOF, based on pillaring of kgm-a or sql-a layers with heterofunctional 3-connected organic building blocks were targeted and constructed to purposely introduce and expose the desired amide or amine functionalities. Interestingly, gas adsorption properties of eea-MOF-4 (1) and eea-MOF-5 (2) showed that by simply altering the nitrogen donor position within the ligand, it is possible to relatively reduce the pore size of the related eea-MOF material and subsequently increase the associated CO2 uptake. The slightly confined pore space in 2, relative to 1, has enabled an enhancement of the pore local charge density and thus the observed relative increase in the CO2 and H2 isosteric heat of adsorption (Qst). In addition, light hydrocarbon adsorption studies revealed that 2 is more selective toward C2H6 and C3H8 over CH4 than 1, as exemplified for C2H6 : CH4 (5 : 95) or C3H8 : CH4 (5 : 95) binary gas mixtures.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  7. Macro-encapsulation of heat storage phase-change materials for use in residential buildings. Final report, September 29, 1976-September 6, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Lane, G.A.; Kott, A.C.; Warner, G.L.; Hartwick, P.B.; Rossow, H.E.

    1978-11-01

    Objectives are to assess the feasibility of macro-encapsulated PCM's for residential solar systems, to develop and evaluate such materials. Encapsulant materials under consideration are multilayer flexible plastic films, steel cans, and plastic bottles. Studies were done in a storage test unit on: CaCl/sub 2/.6H/sub 2/O encapsulated in polyethylene bottles, Mg(NO/sub 3/)/sub 2/6H/sub 2/O in steel cans, Mg(NO/sub 3/)/sub 2/6H/sub 2/O/NH/sub 4/NO/sub 3/ eutectic encapsulated in flexible plastic film packages, NH/sub 4/Br/urea eutectic in polypropylene bottles, and a combined bed of CaCl/sub 2/.6H/sub 2/O and Mg(NO/sub 3/)/sub 2/6H/sub 2/O/NH/sub 4/NO/sub 3/ eutectic. In most cases, charge and discharge capacities in excess of 90% of theory were obtained. Several designs using encapsulated PCM were analyzed, one was selected, and preliminary designs were developed for a demonstration project.

  8. Thermal safety analysis of a dry storage cask for the Korean standard spent fuel - 16159

    International Nuclear Information System (INIS)

    A conceptual dry storage facility, which is based on a commercial dry storage facility, was designed for the Korea standard spent nuclear fuel (SNF) and preliminary thermal safety analysis was performed in this study. To perform the preliminary thermal analysis, a thermal analysis method was proposed. The thermal analysis method consists of 2 parts. By using the method, the surface temperature of the storage canister corresponding to the SNF clad temperature was calculated and the adequate air duct area was decided using the calculation result. The initial temperature of the facility was calculated and the fire condition and half air duct blockage were analyzed. (authors)

  9. Modeling of thermal evolution of near field area around single pit mode nuclear waste canister disposal in soft rocks

    International Nuclear Information System (INIS)

    Soft rocks like argillites/shales are under consideration worldwide as host rock for geological disposal of vitrified as well as spent fuel nuclear waste. The near field around disposed waste canister at 400-500m depth witnesses a complex heat field evolution due to varying thermal characteristics of rocks, coupling with hydraulic processes and varying intensity of heat flux from the canister. Smooth heat dissipation across the rock is desirable to avoid buildup of temperature beyond design limit (100 °C) and resultant micro fracturing due to thermal stresses in the rocks and intervening buffer clay layers. This also causes enhancement of hydraulic conductivity of the rocks, radionuclide transport and greater groundwater ingress towards the canister. Hence heat evolution modeling constitutes an important part of safety assessment of geological disposal facilities

  10. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

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

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.

    1996-11-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Driggers, S.A.

    1994-03-10

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

  14. Critical review of welding technology for canisters for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Nagra is the Swiss national cooperative for the disposal of radioactive waste and is responsible for final disposal of all types of waste produced in Switzerland, which are partitioned into two repository types, one for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate level waste and one for low and intermediate level waste. In the general licences applied for these repositories, documentation has to show that long-term safety can be ensured and that factors for the construction, operation, and closure of the facility have been considered. Nagra has commissioned TWI to carry out a critical review of welding technologies for the sealing of HLW and SF canisters made of carbon steel. In conjunction with a material selection report, the information gained will be used as a preliminary step to provide input to developing design concepts for the canisters. The features to be considered are: a) Suitability of techniques for thickness of weld required; b) Suitability for remote operation, maintenance and set-up; c) Welding speed, weld quality, tolerances and cost; d) Effect of welding process on parent materials properties including microstructure corrosion resistance, distortion and residual stress; e) Potential post-weld treatments to reduce residual stress and enhance corrosion resistance; f) Suitability of inspection techniques for the weld thickness required; g) Impact of welding techniques on the canister design and material selection; h) Critique of emerging technologies which may be suitable in the future. The review of potential welding technologies began with a feasibility study carried out by TWI experts, where the unsuitable processes were rejected. For the remaining processes attention was focused on previous applications for the material and thickness suggested, and especially on safety critical applications such as applied in the nuclear and pressure vessel industry. Once the relevant information was gathered each process was

  15. Creep of OFHC and silver copper at simulated final repository canister-service conditions

    International Nuclear Information System (INIS)

    Results of high-resolution creep rate measurements are described for estimating very long term creep life of copper and silver alloyed copper at room temperature and at stresses approaching the expected service conditions of final repository canisters. The aim was to assess the limiting service stress levels for potential canister wall materials. The 0.1% silver alloyed copper showed minimum creep rates of 10-9 to 10-10 l/h, corresponding to 1 % strain in about 1000 to 10000 years, at room temperature and uniaxial stress level of 50 to 75 MPa. The predicted time to 1 % strain, when extrapolated from literature data, was at least one order of magnitude shorter. From the results of the present work, the 1 % creep life for OFHC copper was at most a few hundreds of years at 50 MPa stress level. The technique developed and used in this work for measuring very low strain rates appears useful for assessing low temperature creep life of practical structures essentially without accelerating the test from the service conditions. (au)

  16. Creep of OFHC and silver copper at simulated final repository canister-service conditions

    International Nuclear Information System (INIS)

    Result of high-resolution creep rate measurements are described for estimating very long term creep life of copper and silver alloyed copper at room temperature and at stresses approaching the expected service conditions of final repository canisters. The aim was to assess the limiting service stress levels for potential canister wall materials. The 0.1 % silver alloyed copper showed minimum creep rates of 10-9 to 10-10 l/h, corresponding to 1 % strain in about 1000 to 10000 years, at room temperature and uniaxial stress level of 50 to 75 MPa. The predicted time to 1 % strain, when extrapolated from literature data, was at least one order of magnitude shorter. From the results of the present work, the 1 % creep life for OFHC copper was at most a few hundreds of years at 50 MPa stress level. The technique developed and used in this work for measuring very low strain rates appears useful for assessing low temperature creep life of practical structures essentially without accelerating the test from the service conditions

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  18. Advantages on dry interim storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, L.S. [Centro Tecnologico da Marinha em Sao Paulo, Av. Professor Lineu Prestes 2468, 05508-900 Sao Paulo (Brazil); Rzyski, B.M. [IPEN/ CNEN-SP, 05508-000 Sao Paulo (Brazil)]. e-mail: romanato@ctmsp.mar.mil.br

    2006-07-01

    When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

  19. Finite element analysis of stresses and deformations occurring in the spent nuclear fuel (SNF) disposal canister deposited in a deep geological repository

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Bo [Department of Mechanical and Aerospace Engineering, Rutgers, The State University of New Jersey, 98 Brett Road, Piscataway, NJ 08854-8058 (United States); School of Aeronautics, Northwestern Polytechnical University of China, Xi’an 710072 (China); Gea, Hae Chang [Department of Mechanical and Aerospace Engineering, Rutgers, The State University of New Jersey, 98 Brett Road, Piscataway, NJ 08854-8058 (United States); Kwon, Young Joo, E-mail: yjkwon@hongik.ac.kr [Department of Mechanical and Design Engineering, Hongik University, 2639 Sejong-ro, Jochiwon, Sejong 339-701 (Korea, Republic of)

    2014-01-15

    Highlights: • The weight reduction design of the PWR SNF disposal canister is presented. • The optimal SNF basket rotation angle to minimize canister weight is sought. • Conventional structural analysis and Kriging method are developed for this purpose. • The optimal SNF basket rotation angle is determined to be 45°. • The canister weight is correspondingly reduced by 16% from 25 tons to 21 tons. -- Abstract: Numerical computer experimental methodologies are investigated for the weight reduction of a spent nuclear fuel (SNF) disposal canister designed to be deposited in a Korean deep geological repository from a pressurized water reactor (PWR). Finite element analyses of stresses and deformations occurring inside the cylindrical canister under the deposited conditions are performed to assess its structural strength at various rotation angles (φ) of the SNF basket. Specifically, the cross sections of four square tube shaped SNF baskets (assemblies) contained in the canister are rotated. Using a conventional structural analysis and a Kriging method, an optimal rotation angle is determined in relation to canister diameter and weight. Both sets of results are in agreement. It was also determined that the computed deformation changes slightly in relation to variances in rotation angle, while the stress incurred inside the cast iron insert of the canister noticeably changes reaching its highest value at φ = 45° while still maintaining safe structural integrity. It is concluded that the diameter of the canister can be reduced from its original design value (102 cm) to 95.8463 cm resulting in a ∼16.0% reduction in canister weight for an optimal rotation angle of 45°.

  20. Thermal performance sensitivity studies in support of material modeling for extended storage of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Suffield, Sarah R.; Fort, James A.; Adkins, Harold E.

    2013-08-15

    The work reported here is an investigation of the sensitivity of component temperatures of a storage system, including fuel cladding temperatures, in response to age-related changes that could degrade the design-basis thermal behavior of the system. Three specific areas of interest were identified for this study. • degradation of the canister backfill gas from pure helium to a mixture of air and helium, resulting from postulated leakage due to stress corrosion cracking (SCC) of canister welds • changes in surface emissivity of system components, resulting from corrosion or other aging mechanisms, which could cause potentially significant changes in temperatures and temperature distributions, due to the effect on thermal radiation exchange between components • changes in fuel and basket temperatures due to changes in fuel assembly position within the basket cells in the canister The purpose of these sensitivity studies is to provide a realistic example of how changes in the physical properties or configuration of the storage system components can affect temperatures and temperature distributions. The magnitudes of these sensitivities can provide guidance for identifying appropriate modeling assumptions for thermal evaluations extending long term storage out beyond 50, 100, 200, and 300 years.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bath, Adrian; Hermansson, Hans-Peter

    2009-08-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-21

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

  3. Indoor radon levels and their relationship with radon exhalation rates from building surface in Hong Kong

    International Nuclear Information System (INIS)

    The indoor radon concentrations of more than 60 sites in Hong Kong is measured using activated charcoal canisters to identify the underlying distribution pattern. The strong relationship between the indoor radon concentrations and the radon exhalation rate from building surface has been investigated. It has been found that the indoor radon comes mainly from radium in building materials, and that the radon concentration depends on the radon exhalation rate from indoor building surface and on the ventilation. It is also asserted that the radioactivity level of building materials used in Hong Kong is increasing

  4. Spent fuel dry storage technology development: thermal evaluation of three adjacent drywells (each containing a 0.6 kW PWR spent fuel assembly)

    International Nuclear Information System (INIS)

    A spent fuel Adjacent Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing three nearly identical pressurized water reactor spent fuel assemblies each having a decay heat level of approximately 0.6 kW. Each fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell storage cell for thermal testing. Each fuel assembly was sealed inside a 14-in. diam, 168-in.-long stainless steel canister and attached to a concrete-filled, 20-in.-diam, 34-in.-long, shield plug. The canister assembly was then placed in a carbon steel drywell liner which had been grouted into a hole drilled in the soil adjacent to E-MAD. The three drywells were located 25 feet apart in a linear array. Thermocouples, provided to measure canister, liner and soil temperatures, were inserted into tubes on the outside of the canister and drywell liner and were attached to plastic pipes which were grouted into holes in the soil. Temperatures from the three drywells and the adjacent soil were recorded throughout the Adjacent Drywell Test. Drywell thermal data showed virtually no thermal interaction between adjacent drywells. However, peak temperatures reached by the three drywells did show a fairly significant difference. Peak canister and drywell liner temperatures were reached in August 1981 for all three drywells. The two previously unused drywells responded similarly with peak canister and liner temperatures reaching 1990F and 1580F, respectively. Comparable peak temperatures for the third drywell which had previously contained spent fuel for nearly 21 months prior to the Adjacent Drywell Test reached 2100F for the canister and 1690F for the drywell liner. This difference is attributed to a decrease in soil thermal conductivity caused by the dryout of soil around the drywell used for previous spent fuel testing

  5. Conceptual design of interim storage facility for CNAI

    International Nuclear Information System (INIS)

    The reduced storage capacity available in the two spent fuel pools of argentine PHWR Atucha-1 power plant, the current plans for extending the reactor operation beyond its design lifetime, and the government decision on Atucha-2 NPP construction ending, have motivated the evaluation of a dry storage option for the interim management of spent fuel assemblies. Two different designs are presently being analyzed by an expert working group, from both technical and economical points of views. Authors are proposing a modular system consisting of an arrangement of reinforced concrete structures into which welded metallic canisters loaded with 37 spent fuel assemblies each stored in horizontal position. The reinforced concrete module is designed to provide the necessary physical protection and biological shielding to the loaded canisters during long-term storage, as well as passive means to remove the spent fuel decay heat by a combination of radiation, conduction and natural air convection. In this works are presented advances in the conceptual designs for a spent nuclear fuel system to Atucha I nuclear power plant. (author)

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

    International Nuclear Information System (INIS)

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

  7. Acceptance of canisters of high-level waste 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. This document discusses the acceptance of canisters of high-level waste by the Federal Waste Management System. 16 refs., 7 figs., 11 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Goldmann, L.H.

    1997-11-10

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars (ed.)

    2006-10-15

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

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

    International Nuclear Information System (INIS)

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

  12. Project JADE. Method and machinery description of equipment for deposition of a canister in a vertical deposition hole

    International Nuclear Information System (INIS)

    A systematic evaluation of different disposal methods has been carried out. The study is named Comparison of Disposal Methods. The evaluation has included a comparison of the technical aspects, safety aspects and costs of alternatives proposed within the so-called KBS-3 method. Three alternatives have been studied and compared: vertical emplacement (KBS-3V), horizontal emplacement (KBS-3H) and emplacement in medium long horizontal holes (MLH). KBS-3V is the reference method adopted in SKB's development and planning work. This report describes eight alternative disposal methods, with variations, and forms a technical basis for the assessment of methods involving vertical disposal (KBS-3V). The alternative of emplacement behind a radiation-shielding screen has been rejected by SKB, as it has been decided that disposal will be carried out with complete radiation shielding around the canister. However, the alternative is considered in the report for the sake of comparison. Based on the applicable technical specifications, the results of fault-effect analyses, radiation protection assessments and flexibility and complexity analyses for the entire disposal process, two methods for vertical emplacement have been identified as the best from a technical point of view: Transport of a horizontally-lying canister which is raised to a vertical position during emplacement. The canister is shielded during transport and the raising movement. Radiation protection can be complete or partial. Transport with a standing canister. Under transport and disposal, the canister is surrounded by a complete radiation shield, which has a telescopic lower part. This principle involves only a few, simple mechanical movements

  13. 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; Gutachten zum Gesuch der ZWILAG Zwischenlager Wuerenlingen AG um Erteilung der Bewilligung fuer Bau und Betrieb des Zentralen Zwischenlagers fuer radioaktive Abfaelle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-15

    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, {alpha}-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 10{sup 20} Bq, including 10{sup 18} Bq of {alpha}-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 10{sup 18} Bq with 10{sup 15} Bq of {alpha}-active nuclides. The maximum thermal power of 250 kW is removed by forced air cooling. Because

  14. New developments in dry spent fuel storage

    International Nuclear Information System (INIS)

    As shown in various new examples, HABOG facility (Netherlands), CERNAVODA (Candu - Romania), KOZLODUY (WWER - Bulgaria), CHERNOBYL ( RMBK - Ukraine), MAYAK (Spent Fuel from submarine and Icebreakers - Russia), recent studies allow to confirm the flexibility and performances of the CASCAD system proposed by SGN, both in safety and operability, for the dry storage of main kinds of spent fuel. The main features are: A multiple containment barrier system: as required by international regulation, 2 independent barriers are provided (tight canister and storage pit); Passive cooling, while the Fuel Assemblies are stored in an inert atmosphere and under conditions of temperature preventing from degradation of rod cladding; Sub-criticality controlled by adequate arrangements in any conditions; Safe facility meeting ICPR 60 Requirements as well as all applicable regulations (including severe weather conditions and earthquake); Safe handling operations; Retrievability of the spent fuel either during storage period or at the end of planned storage period (100 years); Future Decommissioning of the facility facilitated through design optimisation; Construction and operating cost-effectiveness. (author)

  15. Conceptual design for an intermediate dry storage facility for Argentinean Atucha spent fuel

    International Nuclear Information System (INIS)

    Full text: The CNEA (Argentina National Atomic Energy Commission) is planning a new facility for the spent fuel of Atucha I according with the national policy to fulfill the requirement of the National Plan of Radioactive waste management with the lowest cost, having the flexibility to evaluate the fuel back end strategy in a wait and see approach. Spent fuel elements can be stored in concrete for many decades economically and safety as intermediate step, thereby providing adequate time to develop an integrated fuel disposal system, this provides flexibility from the fuel to decay, thus facilitating final disposal with decrease of the decay heat. A centralized storage for the NPP fuel elements (Embalse and Atucha I) with two very different fuel element and different enrichment was not considered, in order to minimize the radioactive waste movement. Nowadays the total life Atucha I spent fuels are in two wet pools, having fuel elements with 28 years old. For Embalse fuel elements type dry vertical concrete silos were successfully implemented for intermediate strategy. An intermediate storage for Atucha I was designed taking into account the following criteria: Assurance the fuel elements integrity for 30 years; Modular build-up to avoid over dimension systems; Low cost radiation shield (concrete and ground); Leak monitoring system for the containment integrity; Possibility to take out the failed containment; Enable the re-encapsulation and the reentry for the fuel containment; Minimize the auxiliary systems with high maintenance cost (passive); Compatible with the national regulatory commission (ARN) regulation with monitoring systems, similar with those implemented in our dry silos at Embalse; Transfer systems and hot cell facility near the pool storage to use its water treatment systems; Minimize secondary waste during wet pool to the intermediate storage. The Atucha I fuel element has 37 fuel rod in circular cluster geometry with an active length of 5,5 meters

  16. High temperature storage loop :

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.

    2013-07-01

    A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650ÀC) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOEs SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

  17. Operational Challenges of Extended Dry Storage of Spent Nuclear Fuel - 12550

    International Nuclear Information System (INIS)

    As a result of the termination of the Yucca Mountain used fuel repository program and a continuing climate of uncertainty in the national policy for nuclear fuel disposition, the likelihood has increased that extended storage, defined as more than 60 years, and subsequent transportation of used nuclear fuel after periods of extended storage may become necessary. Whether at the nation's 104 nuclear energy facilities, or at one or more consolidated interim storage facilities, the operational challenges of extended storage and transportation will depend upon the future US policy for Used Fuel Management and the future Regulatory Framework for EST, both of which should be developed with consideration of their operational impacts. Risk insights into the regulatory framework may conclude that dry storage and transportation operations should focus primarily on ensuring canister integrity. Assurance of cladding integrity may not be beneficial from an overall risk perspective. If assurance of canister integrity becomes more important, then mitigation techniques for potential canister degradation mechanisms will be the primary source of operational focus. If cladding integrity remains as an important focus, then operational challenges to assure it would require much more effort. Fundamental shifts in the approach to design a repository and optimize the back-end of the fuel cycle will need to occur in order to address the realities of the changes that have taken place over the last 30 years. Direct disposal of existing dual purpose storage and transportation casks will be essential to optimizing the back end of the fuel cycle. The federal used fuel management should focus on siting and designing a repository that meets this objective along with the development of CIS, and possibly recycling. An integrated approach to developing US policy and the regulatory framework must consider the potential operational challenges that they would create. Therefore, it should be integral to

  18. Comparison of cask and drywell storage concepts for a monitored retrievable storage/interim storage system

    International Nuclear Information System (INIS)

    The Department of Energy, through its Richland Operations Office is evaluating the feasibility, timing, and cost of providing a federal capability for storing the spent fuel, high-level wastes, and transuranic wastes that DOE may be obligated by law to manage until permanent waste disposal facilities are available. Three concepts utilizing a monitored retrievable storage/interim storage (MRS/IS) facility have been developed and analyzed. The first concept, co-location with a reprocessing plant, has been developed by staff of Allied General Nuclear Services. the second concept, a stand-alone facility, has been developed by staff of the General Atomic Company. The third concept, co-location with a deep geologic repository, has been developed by the Pacific Northwest Laboratory with the assistance of the Westinghouse Hanford Company and Kaiser Engineers. The objectives of this study are: to develop preconceptual designs for MRS/IS facilities: to examine various issues such as transportation of wastes, licensing of the facilities, and environmental concerns associated with operation of such facilities; and to estimate the life-cycle costs of the facilities when operated in response to a set of scenarios that define the quantities and types of waste requiring storage in specific time periods, generally spanning the years 1989 to 2037. Three scenarios are examined to develop estimates of life-cycle costs for the MRS/IS facilities. In the first scenario, the reprocessing plant is placed in service in 1989 and HLW canisters are stored until a repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at intervals as needed to meet the demand. In the second scenario, the reprocessing plants are delayed in starting operations by 10 years, but the repositories open on schedule. In the third scenario, the repositories are delayed 10 years, but the reprocessing plants open on schedule

  19. Comparison of cask and drywell storage concepts for a monitored retrievable storage/interim storage system

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, D.E.

    1982-12-01

    The Department of Energy, through its Richland Operations Office is evaluating the feasibility, timing, and cost of providing a federal capability for storing the spent fuel, high-level wastes, and transuranic wastes that DOE may be obligated by law to manage until permanent waste disposal facilities are available. Three concepts utilizing a monitored retrievable storage/interim storage (MRS/IS) facility have been developed and analyzed. The first concept, co-location with a reprocessing plant, has been developed by staff of Allied General Nuclear Services. the second concept, a stand-alone facility, has been developed by staff of the General Atomic Company. The third concept, co-location with a deep geologic repository, has been developed by the Pacific Northwest Laboratory with the assistance of the Westinghouse Hanford Company and Kaiser Engineers. The objectives of this study are: to develop preconceptual designs for MRS/IS facilities: to examine various issues such as transportation of wastes, licensing of the facilities, and environmental concerns associated with operation of such facilities; and to estimate the life-cycle costs of the facilities when operated in response to a set of scenarios that define the quantities and types of waste requiring storage in specific time periods, generally spanning the years 1989 to 2037. Three scenarios are examined to develop estimates of life-cycle costs for the MRS/IS facilities. In the first scenario, the reprocessing plant is placed in service in 1989 and HLW canisters are stored until a repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at intervals as needed to meet the demand. In the second scenario, the reprocessing plants are delayed in starting operations by 10 years, but the repositories open on schedule. In the third scenario, the repositories are delayed 10 years, but the reprocessing plants open on schedule.

  20. Probe Storage

    NARCIS (Netherlands)

    Gemelli, Marcellino; Abelmann, Leon; Engelen, Johan B.C.; Khatib, Mohammed G.; Koelmans, Wabe W.; Zaboronski, Olog; Campardo, Giovanni; Tiziani, Federico; Laculo, Massimo

    2011-01-01

    This chapter gives an overview of probe-based data storage research over the last three decades, encompassing all aspects of a probe recording system. Following the division found in all mechanically addressed storage systems, the different subsystems (media, read/write heads, positioning, data chan

  1. Grand Challenges facing Storage Systems

    CERN Document Server

    CERN. Geneva

    2004-01-01

    In this talk, we will discuss the future of storage systems. In particular, we will focus on several big challenges which we are facing in storage, such as being able to build, manage and backup really massive storage systems, being able to find information of interest, being able to do long-term archival of data, and so on. We also present ideas and research being done to address these challenges, and provide a perspective on how we expect these challenges to be resolved as we go forward.

  2. A review of materials and corrosion issues regarding canisters for disposal of spent fuel and high-level waste in Opalinus clay

    International Nuclear Information System (INIS)

    The project 'Entsorgungsnachweis' presented by NAGRA to the Swiss Federal Government in December 2002 assessed the feasibility of disposal of spent fuel (SF), vitrified high level waste (HLW) from reprocessing and long-lived intermediate level waste in an Opalinus Clay repository site in Northern Switzerland. NAGRA proposed the use of carbon steel canisters for disposal of SF/HLW and it also put forward an alternative concept of copper canisters with cast iron insert. In its reply the Federal Government acknowledged that NAGRA had successfully demonstrated the technical feasibility of disposal of SF/HLW. However, some of its experts raised a number of questions related to the choice of steel as canister material. Among others, it was questioned whether hydrogen formed by corrosion of steel in contact with saturated bentonite might adversely affect the barrier function of the Opalinus clay. It was also recommended that alternative canister materials and/or design concepts should be evaluated. To deal with these concerns NAGRA convened an international group of experts, the Canister Materials Review Board (CMRB), who were to review the existing information on canister materials that could be suitable for the proposed repository environment. Based on present knowledge of materials science, the CMRB was to recommend to NAGRA the most suitable material(s) for meeting the performance requirements for SF/HLW canisters. Specifically, the CMRB was to consider corrosion, including hydrogen generation, and stress-assisted failure processes that could affect the integrity and projected life time of SF/HLW canisters or impede the functioning of geological barriers while keeping in mind the overall feasibility of manufacturing, sealing and inspecting the canisters. The CMRB was further asked to identify the needs and provide advice for further studies by NAGRA on the long term performance and safety of SF/HLW canisters in the Swiss repository concept. For the assessment of the

  3. Spent-fuel dry-storage testing at E-MAD (March 1978-March 1982)

    International Nuclear Information System (INIS)

    From March 1978 through March 1982, spent fuel dry storage tests were conducted at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site to confirm that commercial reactor spent fuel could be encapsulated and passively stored in one or more interim dry storage cell concepts. These tests were: electrically heated drywell, isolated and adjacent drywell, concrete silo, fuel assembly internal temperature measurement, and air-cooled vault. This document presents the test data and results as well as results from supporting test operations (spent fuel calorimetry and canister gas sampling)

  4. Description of a Multipurpose Processing and Storage Complex for the Hanford Site`s radioactive material

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, D.H.; Wolfe, B.A. [Westinghouse Hanford Co., Richland, WA (United States); Hoertkorn, T.R. [USDOE Richland Operations Office, WA (United States)

    1993-05-01

    The mission of the US Department of Energy`s (DOE) Hanford Site has changed from defense nuclear materials production to that of waste management/disposal and environmental restoration. ne Multipurpose Processing and Storage Complex (MPSC) is being designed to process discarded waste tank internal hardware contaminated with mixed wastes, failed melters from the vitrification plant, and other Hanford Site high-level solid waste. The MPSC also will provide interim storage of other radioactive materials (irradiated fuel, canisters of vitrified high-level waste [HLW], special nuclear material [SNM], and other designated radioactive materials).

  5. Spent-fuel dry-storage testing at E-MAD (March 1978-March 1982)

    Energy Technology Data Exchange (ETDEWEB)

    Unterzuber, R.; Milnes, R.D.; Marinkovich, B.A.; Kubancsek, G.M.

    1982-09-01

    From March 1978 through March 1982, spent fuel dry storage tests were conducted at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site to confirm that commercial reactor spent fuel could be encapsulated and passively stored in one or more interim dry storage cell concepts. These tests were: electrically heated drywell, isolated and adjacent drywell, concrete silo, fuel assembly internal temperature measurement, and air-cooled vault. This document presents the test data and results as well as results from supporting test operations (spent fuel calorimetry and canister gas sampling).

  6. Description of a Multipurpose Processing and Storage Complex for the Hanford Site's radioactive material

    International Nuclear Information System (INIS)

    The mission of the US Department of Energy's (DOE) Hanford Site has changed from defense nuclear materials production to that of waste management/disposal and environmental restoration. ne Multipurpose Processing and Storage Complex (MPSC) is being designed to process discarded waste tank internal hardware contaminated with mixed wastes, failed melters from the vitrification plant, and other Hanford Site high-level solid waste. The MPSC also will provide interim storage of other radioactive materials (irradiated fuel, canisters of vitrified high-level waste [HLW], special nuclear material [SNM], and other designated radioactive materials)

  7. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  8. Pilot study on diffusive ground heat storage in the SUVA 'D4' building complex in Root, Switzerland; Etude pilote pour le stockage diffusif des batiments du centre D4 de la SUVA a Root, Lucerne. Analyse de 2 tests de reponse geothermique et integration du stockage diffusif dans le systeme

    Energy Technology Data Exchange (ETDEWEB)

    Pahud, D.

    2001-07-01

    This report for the Swiss Federal Office of Energy (SFOE) presents, in a first part, the results of two thermal response tests made on a geothermal heat storage system that forms part of a heating and cooling system for a complex of buildings comprising the 'D4' centre of the Swiss Accident Insurance Institution (SUVA) in Root, near Lucerne, Switzerland. These in situ response tests on two borehole heat exchangers confirmed the laboratory measurements, indicating that no significant ground water flow is present. In a second part the heat storage dimensioning and the heating/cooling system evaluation by means of computerized simulation are described, which used simulated data for the heating and cooling requirements of the new buildings and the ground heat storage parameters. The PILESIM thermal simulation tool is described, a computer code used to simulate heating and cooling systems that use heat-exchanger piles or borehole heat exchangers. The evaluations made on three system concepts are discussed that permitted the optimal design of the diffusive ground storage system. The construction of the actual system is also described.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

  10. Monitored Retrievable Storage System Requirements Document. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-03-01

    This Monitored Retrievable Storage System Requirements Document (MRS-SRD) describes the functions to be performed and technical requirements for a Monitored Retrievable Storage (MRS) facility subelement and the On-Site Transfer and Storage (OSTS) subelement. The MRS facility subelement provides for temporary storage, at a Civilian Radioactive Waste Management System (CRWMS) operated site, of spent nuclear fuel (SNF) contained in an NRC-approved Multi-Purpose Canister (MPC) storage mode, or other NRC-approved storage modes. The OSTS subelement provides for transfer and storage, at Purchaser sites, of spent nuclear fuel (SNF) contained in MPCs. Both the MRS facility subelement and the OSTS subelement are in support of the CRWMS. The purpose of the MRS-SRD is to define the top-level requirements for the development of the MRS facility and the OSTS. These requirements include design, operation, and decommissioning requirements to the extent they impact on the physical development of the MRS facility and the OSTS. The document also presents an overall description of the MRS facility and the OSTS, their functions (derived by extending the functional analysis documented by the Physical System Requirements (PSR) Store Waste Document), their segments, and the requirements allocated to the segments. In addition, the top-level interface requirements of the MRS facility and the OSTS are included. As such, the MRS-SRD provides the technical baseline for the MRS Safety Analysis Report (SAR) design and the OSTS Safety Analysis Report design.

  11. Factors Affecting the Estimation of Indoor Radon Using Passive Activated Charcoal Canisters

    Science.gov (United States)

    Scarpitta, Salvatore Charles

    1990-01-01

    Adsorption and desorption studies of 20 activated charcoals were conducted in a monolayer and a packed bed utilizing tracer gases. Kinetic studies, using xenon-133, demonstrate the existence of a two-compartment micropore volume with entrance capillaries which together determine the response characteristics of the charcoal to external concentration gradients of tracer gases. This new two -compartment model adequately describes the adsorption and desorption dynamics of radon in the presence of water vapor. Measurements with charcoal exposed to water vapor and Rn-222 in a monolayer and packed bed for exposure intervals of 1-7 days demonstrate that the uptake rate and total quantity of adsorbed Rn-222 are highly dependent upon the amount of water adsorbed. The effect of CO_2 on radon adsorption is small in any charcoal. The measured effective diffusion coefficient of radon in a packed bed of a peat based charcoal at 15% humidity and 25^circC is 7.97 times 10^{-6} cm^2/s. Condensed water vapor in the entrance capillaries reduces the effective pore radius, increasing the diffusion half-time, both into and out of the charcoal. The amount of adsorbed water per gram of charcoal required to block the entrance capillaries varies with the charcoal type. The proposed term for this quantity is the "break-point". A two-stage diffusion barrier charcoal monitor with a long diffusion path length was developed. This design inhibits passive airflow while maintaining the amount of adsorbed water vapor in the primary charcoal adsorbent below the break-point. Water removal at the entry port allows for longer exposure times improving the integrating capability necessary for indoor exposure assessment. The long diffusion path length increases the integration time -constant for radon adsorption normally 24 hours for conventional open-faced canisters to 50 hours for the improved canister. The increased integration time-constant allows for a 7 day sample to be measured at 70% humidity and 23

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-15

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

  13. Brine: a computer program to compute brine migration adjacent to a nuclear waste canister in a salt repository

    International Nuclear Information System (INIS)

    This report presents a mathematical model used to predict brine migration toward a nuclear waste canister in a bedded salt repository. The mathematical model is implemented in a computer program called BRINE. The program is written in FORTRAN and executes in the batch mode on a CDC 7600. A description of the program input requirements and output available is included. Samples of input and output are given

  14. Monitoring plan for routine organic air emissions at the Radioactive Waste Management Complex Waste Storage Facilities

    International Nuclear Information System (INIS)

    This monitoring plan provides the information necessary to perform routine organic air emissions monitoring at the Waste Storage Facilities located at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The Waste Storage Facilities include both the Type I and II Waste Storage Modules. The plan implements a dual method approach where two dissimilar analytical methodologies, Open-Path Fourier Transform Infrared Spectroscopy (OP-FTIR) and ancillary SUMMA reg-sign canister sampling, following the US Environmental Protection Agency (EPA) analytical method TO-14, will be used to provide qualitative and quantitative volatile organic concentration data. The Open-Path Fourier Transform Infrared Spectroscopy will provide in situ, real time monitoring of volatile organic compound concentrations in the ambient air of the Waste Storage Facilities. To supplement the OP-FTIR data, air samples will be collected using SUMMA reg-sign, passivated, stainless steel canisters, following the EPA Method TO-14. These samples will be analyzed for volatile organic compounds with gas chromatograph/mass spectrometry analysis. The sampling strategy, procedures, and schedules are included in this monitoring plan. The development of this monitoring plan is driven by regulatory compliance to the Resource Conservation and Recovery Act, State of Idaho Toxic Air Pollutant increments, Occupational Safety and Health Administration. The various state and federal regulations address the characterization of the volatile organic compounds and the resultant ambient air emissions that may originate from facilities involved in industrial production and/or waste management activities

  15. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Special study No. 3. Waste retrieval from backfilled regions

    International Nuclear Information System (INIS)

    Methods and costs were studied for delayed canister retrieval from rooms that had been backfilled immediately after canister storage. The effects of this method of storage on mine geometry, thermal and rock mechanics environments, mine development and operations, mine ventilation, time schedule, retrieval machinery and safety were investigated. Salt and air temperatures were determined. Pillar width, number of rooms, extraction ratio, tonnages of mined salt, and salt handling and hoisting requirements were calculated. The required changes in mining equipment were established. Salt handling and elapsed time schedules were developed. Ventilation requirements - size and number of shafts, size the arrangement of airways, number of stacks, and size and number of fans were then calculated. The development sequence of these facilities was established. Canister retrieval problems were analyzed for canisters stuck in the hole as well as free. Retrieval methods and machinery were studied and are described. Safety with respect to both radiation and room collapse was studied and compared with CDR safety conditions. The effects of a reduced themal loading of 30 KW/acre on temperatures, room closure, mine layout, ventilation and ground control were studied and reported. A cost estimate was prepared, giving cost differentials between the base CDR costs and Special Study No. 3. Two appendices are included. The first contains nine Heat Transfer memoranda that state the thermal basis of this study. The second appendix provides a detailed operating time analysis of the retrieval machinery

  16. Industrial Storage Technology Applied to Library Requirements.

    Science.gov (United States)

    Kountz, John

    1987-01-01

    Comparison of conventional book stacks, moving aisle book stacks, and industrial storage techniques shows that the industrial technique is the least expensive to build, maintain, and operate. Characteristics of automated storage and retrieval systems are discussed, and formulas for calculating the size and costs of a library system are presented.…

  17. The leading of Titanium on corrosion resistance of AISI 321 stainless steel as material for nuclear waste canister

    International Nuclear Information System (INIS)

    Ultimate disposal of spent fuel or high level waste on underground, needs canister which has high corrosion resistance. The AISI 321 stainless steel which contains titanium as spent fuel or high level waste canister has been studied. The titanium content in alloy would retard the formation of Cr23C6 precipitate, so the corrosion attack will be avoided. The experiment was conducted by heating specimens at temperature of 700oC for 2 hours in which the analysis was performed by optical microscope, SEM and EDS. The analysis results showed that at the experiment temperature was undetected the Cr23C6 precipitate, although the TiC precipitate was formed with average diameter of 4.70 μm with the Ti content on the TiC and matrix were 96.20% and 0.925% of weight respectively. The fact that this material has a high corrosion resistance, so the use as spent fuel or high level waste canister will be sufficiently safe

  18. Demonstrating compliance with the waste acceptance preliminary specifications on foreign materials within DWPF canistered waste forms (U)

    International Nuclear Information System (INIS)

    The Defense Waste Processing Facility (DWPF) will employ a waste acceptance program based on the Waste Acceptance Preliminary Specifications (WAPS). These specifications require, among other criteria, that the canistered waste form contain no free liquids, free gases, organics, or explosives. Of particular importance is the absence of liquid water. This paper summarizes efforts and discusses experiments at the Savannah River Site for demonstrating compliance with the foreign materials specifications of the WAPS. Existing data, already in the literature, is being combined with the results of new experiments. For the volatility of the waste glass, documented work is combined with new results of thermogravimetric analysis experiments on simulated waste glass samples produced during scale glass melter campaigns. The volatility of these glass samples provides evidence that no free liquids, free gases, organics, or explosives are released upon heating the waste glass to its glass transition temperature. To show compliance of the absence of liquid water, documented work is being combined with the results of new experiments involving measurement of the internal gas pressure, the composition of the gas within the canisters, and the relative humidity of sealed, canistered waste forms produced during large-scale glass melter runs and the upcoming cold runs of the DWPF. (orig.)

  19. Pemakaian Thermal Storage pada Sistem Pengkondisi Udara

    Directory of Open Access Journals (Sweden)

    Herry Sunandar

    1999-01-01

    Full Text Available Air conditioning system consume 70 - 80 % of a building's energy requirement. Application of ice storage system can regulate electric load consumption so that its peak load can be controlled. Abstract in Bahasa Indonesia : Sistem pengkondisian udara memakai energi 70 - 80 % dari seluruh energy sebuah gedung. Penerapan ice storage pada sistem tersebut dapat mengatur pemakaian beban listrik sehingga pemakaian beban listrik pada beban puncak dapat dikontrol. Kata kunci : ice storage, chiller, sistem pengkondisi udara

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

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R.; Enos, David George

    2014-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    J.R. Massari

    1995-10-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  5. Nano-storage wires.

    Science.gov (United States)

    Lee, Dong Jun; Kim, Eunji; Kim, Daesan; Park, Juhun; Hong, Seunghun

    2013-08-27

    We report the development of "nano-storage wires" (NSWs), which can store chemical species and release them at a desired moment via external electrical stimuli. Here, using the electrodeposition process through an anodized aluminum oxide template, we fabricated multisegmented nanowires composed of a polypyrrole segment containing adenosine triphosphate (ATP) molecules, a ferromagnetic nickel segment, and a conductive gold segment. Upon the application of a negative bias voltage, the NSWs released ATP molecules for the control of motor protein activities. Furthermore, NSWs can be printed onto various substrates including flexible or three-dimensional structured substrates by direct writing or magnetic manipulation strategies to build versatile chemical storage devices. Since our strategy provides a means to store and release chemical species in a controlled manner, it should open up various applications such as drug delivery systems and biochips for the controlled release of chemicals.

  6. ANALYSIS OF SLUDGE BATCH 4 (MACROBATCH 5) FOR CANISTER S02902 AND SLUDGE BATCH 5 (MACROBATCH 6) FOR CANISTER S03317 DWPF POUR STREAM GLASS SAMPLES

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M.; Bibler, N.

    2010-10-04

    The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 4 (SB4), Macrobatch 5 (MB5) on May 29, 2007. Sludge Batch 4 was a blend of the heel of Tank 40 from Sludge Batch 3 (SB3) and SB4 material qualified in Tank 51. On November 28, 2008, DWPF began processing Sludge Batch 5 (SB5) from Tank 40 which is a blend of the heel of Tank 40 from SB4, SB5 material qualified in Tank 51 and H-Canyon Pu and Np transfers. SB4 was processed using Frit 510 and SB5 used Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. During the processing of SB4 and SB5, glass samples were obtained during the pouring of canisters S02902 and S03317, respectively. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed (durability, chemical and radionuclide composition). The following observations and conclusions are drawn from the analytical results provided in this report: (1) The sum of the oxides for the chemical composition of both the SB4 and SB5 pour stream glasses is within the Product Composition Control System (PCCS) acceptance limits (95 {le} sum of oxides {le} 105). (2) The calculated Sludge Dilution Factor (SDF) for SB4 is 2.52. The measured radionuclide content is in good agreement with the calculated values from the dried sludge results from the SB4 Waste Acceptance Production Specification (WAPS) sample (References 1 and 19). (3) The calculated SDF for SB5 is 2.60. The measured radionuclide content is in good agreement with the calculated values from the dried sludge results from the SB5 WAPS sample (References 2 and 20). (4) Scanning Electron Microscopy (SEM) analysis shows there are noble metal inclusions, primarily ruthenium, present in both pour stream samples. (5) The Product

  7. Feasibility study of recycle material storage facility in the flexible fuel cycle system during transition from LWR cycle to FBR cycle

    International Nuclear Information System (INIS)

    In order to establish the future ideal FBR fuel cycle system from the current LWR fuel cycle, we proposed the Flexible Fuel Cycle Initiative (FFCI) as a suitable transition system. The FFCI removes about 90% of the U present in LWR spent fuel and the residual material, designated as recycle material (RM), is temporarily stored in a facility, such as the vitrified high-level waste (HLW) storage facility. Decay heat of the RM is removed with natural convection of air. In this paper, feasibility of the RM storage facility was studied from viewpoints of heat removal performance and critical safety. Maximum temperature of the RM and the containment pipe were evaluated for the variable density value which depends on the RM filling ratio. Because the heat density of the RM is larger than that of vitrified HLW, we made the diameter of the canister 15cm which is about one third the diameter of a vitrified HLW canister. To improve economy, we evaluated the safety of the storage facility in which three canisters surrounded by an outer pipe identical to the air pipe used in the HLW storage facility. Each canister is surrounded by an containment pipe and an pipe. The points of the feasibility study for the RM storage facility were summarized as follows. 1) The thermal analysis of the storage facility was done by regarding RM as a granule material and basically consisting of uranium oxide. Temperature at the center of the RM was evaluated with various thermal conductivity models in relation to RM filling ratio. 2) Temperature of the RM storage facility was also evaluated preliminarily with thermal properties of simulated oxide RM. Temperature at the center of a canister slightly exceeded a limit value of 1000degC for the RM produced in an air atmosphere. But it can be de reduced within the limit for the RM produced in a helium atmosphere. 3) Critical safety was also investigated for RM storage and it was found that sub-criticality could be achieved even in the hypothetical

  8. Monitored retrievable storage submission to Congress: Volume 2, Environmental assessment for a monitored retrievable storage facility

    International Nuclear Information System (INIS)

    This Environmental Assessment (EA) supports the DOE proposal to Congress to construct and operate a facility for monitored retrievable storage (MRS) of spent fuel at a site on the Clinch River in the Roane County portion of Oak Ridge, Tennessee. The first part of this document is an assessment of the value of, need for, and feasibility of an MRS facility as an integral component of the waste management system. The second part is an assessment and comparison of the potential environmental impacts projected for each of six site-design combinations. The MRS facility would be centrally located with respect to existing reactors, and would receive and canister spent fuel in preparation for shipment to and disposal in a geologic repository. 207 refs., 57 figs., 132 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  11. Energy Storage

    Energy Technology Data Exchange (ETDEWEB)

    Mukundan, Rangachary [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-30

    Energy storage technology is critical if the U.S. is to achieve more than 25% penetration of renewable electrical energy, given the intermittency of wind and solar. Energy density is a critical parameter in the economic viability of any energy storage system with liquid fuels being 10 to 100 times better than batteries. However, the economical conversion of electricity to fuel still presents significant technical challenges. This project addressed these challenges by focusing on a specific approach: efficient processes to convert electricity, water and nitrogen to ammonia. Ammonia has many attributes that make it the ideal energy storage compound. The feed stocks are plentiful, ammonia is easily liquefied and routinely stored in large volumes in cheap containers, and it has exceptional energy density for grid scale electrical energy storage. Ammonia can be oxidized efficiently in fuel cells or advanced Carnot cycle engines yielding water and nitrogen as end products. Because of the high energy density and low reactivity of ammonia, the capital cost for grid storage will be lower than any other storage application. This project developed the theoretical foundations of N2 catalysis on specific catalysts and provided for the first time experimental evidence for activation of Mo 2N based catalysts. Theory also revealed that the N atom adsorbed in the bridging position between two metal atoms is the critical step for catalysis. Simple electrochemical ammonia production reactors were designed and built in this project using two novel electrolyte systems. The first one demonstrated the use of ionic liquid electrolytes at room temperature and the second the use of pyrophosphate based electrolytes at intermediate temperatures (200 – 300 ºC). The mechanism of high proton conduction in the pyrophosphate materials was found to be associated with a polyphosphate second phase contrary to literature claims and ammonia production rates as high as 5X 10

  12. Long-Term Dry Storage of High Burn-Up Spent Pressurized Water Reactor (PWR) Fuel in TAD (Transportation, Aging, and Disposal) Containers

    International Nuclear Information System (INIS)

    A TAD canister, in conjunction with specially-designed over-packs can accomplish the functions of transportation, aging, and disposal (TAD) in the management of spent nuclear fuel (SNF). Industrial dry cask systems currently available for SNF are licensed for storage-only or for dual-purpose (i.e., storage and transportation). By extending the function to include the indefinite storage and perhaps, eventual geologic disposal, the TAD canister would have to be designed to enhance, among others, corrosion resistance, thermal stability, and criticality-safety control. This investigative paper introduces the use of these advanced iron-based, corrosion-resistant materials for SNF transportation, aging, and disposal.The objective of this investigative project is to explore the interest that KAERI would research and develop its specific SAM coating materials for the TAD canisters to satisfy the requirements of corrosion-resistance, thermal stability, and criticality-controls for long-term dry storage of high burn-up spent PWR fuel

  13. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  14. Safety research activities for Japanese regulations of spent fuel interim storage facilities

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) carries out (a) preparation of technical documents, (b) technical evaluations of standards (prepared by academic societies), etc. and (c) other R and D activities, to support Nuclear and Industrial Safety Agency (NISA: which prepares necessary regulations for Spent Fuel Interim Storage Facilities). In 2010 fiscal year, JNES completed technical evaluation of the standard (prepared by Atomic Energy Society of Japan) used for the storage facility (dual purpose cask system) being constructed in Mutsu-City and R and D for UT test of welded canister lids which is required for concrete cask storage facilities. And also, JNES is preparing dynamic test of spent fuel to examine the integrity of spent fuel at cask drop accidents and PWR spent fuel storage test to prove long term integrity of spent fuel and cask itself. The results of these tests will be reported in 2011 and 2012 fiscal year. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-09-15

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-01-01

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

  19. Building America

    Energy Technology Data Exchange (ETDEWEB)

    Brad Oberg

    2010-12-31

    IBACOS researched the constructability and viability issues of using high performance windows as one component of a larger approach to building houses that achieve the Building America 70% energy savings target.

  20. A Preliminary Evaluation of Using Fill Materials to Stabilize Used Nuclear Fuel During Storage and Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Lahti, Erik A.; Richmond, David J.

    2012-08-01

    This report contains a preliminary evaluation of potential fill materials that could be used to fill void spaces in and around used nuclear fuel contained in dry storage canisters in order to stabilize the geometry and mechanical structure of the used nuclear fuel during extended storage and transportation after extended storage. Previous work is summarized, conceptual descriptions of how canisters might be filled were developed, and requirements for potential fill materials were developed. Elements of the requirements included criticality avoidance, heat transfer or thermodynamic properties, homogeneity and rheological properties, retrievability, material availability and cost, weight and radiation shielding, and operational considerations. Potential fill materials were grouped into 5 categories and their properties, advantages, disadvantages, and requirements for future testing were discussed. The categories were molten materials, which included molten metals and paraffin; particulates and beads; resins; foams; and grout. Based on this analysis, further development of fill materials to stabilize used nuclear fuel during storage and transportation is not recommended unless options such as showing that the fuel remains intact or canning of used nuclear fuel do not prove to be feasible.