Sample records for dry cask storage

  1. Dry Cask Storage Study Feb 1989

    Broader source: Energy.gov [DOE]

    This report on the use of dry-cask-storage technologies at the sites of civilian nuclear power reactors has been prepared by the U.S. Department of Energy (DOE} in response to the requirements of...

  2. Inspection of Used Fuel Dry Storage Casks

    SciTech Connect (OSTI)

    Dennis C. Kunerth; Tim McJunkin; Mark McKay; Sasan Bakhtiari

    2012-09-01T23:59:59.000Z

    ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) regulates the storage of used nuclear fuel, which is now and will be increasingly placed in dry storage systems. Since a final disposition pathway is not defined, the fuel is expected to be maintained in dry storage well beyond the time frame originally intended. Due to knowledge gaps regarding the viability of current dry storage systems for long term use, efforts are underway to acquire the technical knowledge and tools required to understand the issues and verify the integrity of the dry storage system components. This report summarizes the initial efforts performed by researchers at Idaho National Laboratory and Argonne National Laboratory to identify and evaluate approaches to in-situ inspection dry storage casks. This task is complicated by the design of the current storage systems that severely restrict access to the casks.

  3. Interim storage cask (ISC), a concrete and steel dry storage cask

    SciTech Connect (OSTI)

    Grenier, R.M.; Koploy, M.A. [General Atomics, San Diego, CA (United States)

    1995-12-31T23:59:59.000Z

    General Atomics (GA) has designed and is currently fabricating the Interim Storage Cask (ISC) for Westinghouse Hanford Company (WHC). The ISC is a dry storage cask that will safely store a Core Component Container (CCC) with Fast Flux Test Facility (FFTF) spent fuel assemblies or fuel pin containers for a period of up to 50 years at the US Department of Energy (DOE) Hanford site. The cask may also be used to transfer the fuel to different areas within the Hanford site. The ISC is designed to stringent criteria from both 10CFR71 and 10CFR72 for safe storage and on-site transportation of FFTF spent fuel and fuel pin containers. The cask design uses a combination of steel and concrete materials to achieve a cost-effective means of storing spent fuel. The casks will be extensively tested before use to verify that the design and construction meet the design requirements.

  4. Standard review plan for dry cask storage systems. Final report

    SciTech Connect (OSTI)

    NONE

    1997-01-01T23:59:59.000Z

    The Standard Review Plan (SRP) For Dry Cask Storage Systems provides guidance to the Nuclear Regulatory Commission staff in the Spent Fuel Project Office for performing safety reviews of dry cask storage systems. The SRP is intended to ensure the quality and uniformity of the staff reviews, present a basis for the review scope, and clarification of the regulatory requirements. Part 72, Subpart B generally specifies the information needed in a license application for the independent storage of spent nuclear fuel and high level radioactive waste. Regulatory Guide 3.61 {open_quotes}Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask{close_quotes} contains an outline of the specific information required by the staff. The SRP is divided into 14 sections which reflect the standard application format. Regulatory requirements, staff positions, industry codes and standards, acceptance criteria, and other information are discussed.

  5. Viability of Existing INL Facilities for Dry Storage Cask Handling

    SciTech Connect (OSTI)

    Randy Bohachek; Charles Park; Bruce Wallace; Phil Winston; Steve Marschman

    2013-04-01T23:59:59.000Z

    This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

  6. Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination

    SciTech Connect (OSTI)

    Bare, Walter Claude; Ebner, Matthias Anthony; Torgerson, Laurence Dale

    2001-08-01T23:59:59.000Z

    This report documents visual examination and testing conducted in 1999 and early 2000 at the Idaho National Engineering and Environmental Laboratory (INEEL) on a Gesellschaft für Nuklear Service (GNS) CASTOR V/21 pressurized water reactor (PWR) spent fuel dry storage cask. The purpose of the examination and testing is to develop a technical basis for renewal of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level waste at independent spent fuel storage installation sites. The examination and testing was conducted to assess the condition of the cask internal and external surfaces, cask contents consisting of 21 Westinghouse PWR spent fuel assemblies from Dominion’s (formerly named Virginia Power) Surry Power Station and cask concrete storage pad. The assemblies have been continuously stored in the CASTOR cask since 1985. Cask exterior surface and selected fuel assembly temperatures, and cask surface gamma and neutron dose rates were measured. Cask external/internal surfaces, fuel basket components including accessible weldments, fuel assembly exteriors, and primary lid seals were visually examined. Selected fuel rods were removed from one fuel assembly, visually examined, and then shipped to Argonne National Laboratory for nondestructive, destructive, and mechanical examination. Cask interior crud samples and helium cover gas samples were collected and analyzed. The results of the examination and testing indicate the concrete storage pad, CASTOR V/21 cask, and cask contents exhibited sound structural and seal integrity and that long-term storage has not caused detectable degradation of the spent fuel cladding or the release of gaseous fission products between 1985 and 1999.

  7. Seismic Behavior of Spent Fuel Dry Cask Storage Systems

    SciTech Connect (OSTI)

    Shaukat, Syed K. [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States); Luk, Vincent K. [Sandia National Laboratories, PO Box 5800. Albuquerque, New Mexico 87185-0744 (United States)

    2002-07-01T23:59:59.000Z

    The U. S. Nuclear Regulatory Commission (NRC) is conducting a research program to investigate technical issues concerning the dry cask storage systems of spent nuclear fuel by conducting confirmatory research for establishing criteria and review guidelines for the seismic behavior of these systems. The program focuses on developing 3-D finite element analysis models that address the dynamic coupling of a module/cask, a flexible concrete pad, and an underlying soil/rock foundation, in particular, the soil-structure-interaction. Parametric analyses of the coupled models are performed to include variations in module/cask geometry, site seismicity, underlying soil properties, and cask/pad interface friction. The analyses performed include: 1) a rectangular dry cask module typical of Transnuclear West design at a site in Western USA where high seismicity is expected; 2) a cylindrical dry cask typical of Holtec design at a site in Eastern USA where low seismicity is expected; and 3) a cylindrical dry cask typical of Holtec design at a site in Western USA with medium high seismicity. The paper includes assumptions made in seismic analyses performed, results, and conclusions. (authors)

  8. Adapting Dry Cask Storage for Aging at a Geologic Repository

    SciTech Connect (OSTI)

    C. Sanders; D. Kimball

    2005-08-02T23:59:59.000Z

    A Spent Nuclear Fuel (SNF) Aging System is a crucial part of operations at the proposed Yucca Mountain repository in the United States. Incoming commercial SNF that does not meet thermal limits for emplacement will be aged on outdoor pads. U.S. Department of Energy SNF will also be managed using the Aging System. Proposed site-specific designs for the Aging System are closely based upon designs for existing dry cask storage (DCS) systems. This paper evaluates the applicability of existing DCS systems for use in the SNF Aging System at Yucca Mountain. The most important difference between existing DCS facilities and the Yucca Mountain facility is the required capacity. Existing DCS facilities typically have less than 50 casks. The current design for the aging pad at Yucca Mountain calls for a capacity of over 2,000 casks (20,000 MTHM) [1]. This unprecedented number of casks poses some unique problems. The response of DCS systems to off-normal and accident conditions needs to be re-evaluated for multiple storage casks. Dose calculations become more complicated, since doses from multiple or very long arrays of casks can dramatically increase the total boundary dose. For occupational doses, the geometry of the cask arrays and the order of loading casks must be carefully considered in order to meet ALARA goals during cask retrieval. Due to the large area of the aging pad, skyshine must also be included when calculating public and worker doses. The expected length of aging will also necessitate some design adjustments. Under 10 CFR 72.236, DCS systems are initially certified for a period of 20 years [2]. Although the Yucca Mountain facility is not intended to be a storage facility under 10 CFR 72, the operational life of the SNF Aging System is 50 years [1]. Any cask system selected for use in aging will have to be qualified to this design lifetime. These considerations are examined, and a summary is provided of the adaptations that must be made in order to use DCS technologies successfully at a geologic repository.

  9. The Feasibility of Cask "Fingerprinting" as a Spent-Fuel, Dry-Storage Cask Safeguards Technique

    SciTech Connect (OSTI)

    Ziock, K P; Vanier, P; Forman, L; Caffrey, G; Wharton, J; Lebrun, A

    2005-07-27T23:59:59.000Z

    This report documents a week-long measurement campaign conducted on six, dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. A gamma-ray imager, a thermal-neutron imager and a germanium spectrometer were used to collect data on the casks. The campaign was conducted to examine the feasibility of using the cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the germanium spectrometer and differences between thermal neutron images and neutron-dose meters, this result is thought to be due to the limitations of the extant imagers used, rather than of the basic concept. Results indicate that measurements with improved imagers could contain significantly more information. Follow-on measurements with new imagers either currently available as laboratory prototypes or under development are recommended.

  10. Managing Aging Effects on Dry Cask Storage

    E-Print Network [OSTI]

    Kemner, Ken

    Because there is currently no designated disposal site for used nuclear fuel in the United States transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance

  11. Regulators Experiences in Licensing and Inspection of Dry Cask Storage Facilities

    SciTech Connect (OSTI)

    Baggett, S.; Brach, E.W. [Spent Fuel Project Office, U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2006-07-01T23:59:59.000Z

    The United States Nuclear Regulatory Commission (NRC), through the combination of a rigorous licensing and inspection program, ensures the safety and security of dry cask storage. NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site-specific licensing and general licensing. In July 1986, the NRC issued the first site-specific license to the Surry Nuclear Power Plant in Virginia authorizing the interim storage of spent fuel in a dry storage cask configuration. Today, there are over 30 ISFSIs currently licensed by the NRC with over 700 loaded dry casks. Current projections identify over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. This paper discusses the NRC licensing and inspection experiences. (authors)

  12. Radiation Imaging of Dry-Storage Casks for Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Ziock, K; Caffrey, G; Lebrun, A; Forman, L; Vanier, P; Wharton, J

    2005-11-08T23:59:59.000Z

    The authors report the results of a measurement campaign conducted on six dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. a gamma-ray imager, a thermal-neutron imager and a Ge-spectrometer were used to collect data. The campaign was conducted to examine the feasibility of using cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the Ge-spectrometer and differences between thermal neutron images and neutron-dose meters, this result is attributed to the limitations of the extant imagers used, rather than of the basic concept.

  13. Compton Dry-Cask Imaging System

    ScienceCinema (OSTI)

    None

    2013-05-28T23:59:59.000Z

    The Compton-Dry Cask Imaging Scanner is a system that verifies and documents the presence of spent nuclear fuel rods in dry-cask storage and determines their isotopic composition without moving or opening the cask. For more information about this project, visit http://www.inl.gov/rd100/2011/compton-dry-cask-imaging-system/

  14. Compton Dry-Cask Imaging System

    SciTech Connect (OSTI)

    None

    2011-01-01T23:59:59.000Z

    The Compton-Dry Cask Imaging Scanner is a system that verifies and documents the presence of spent nuclear fuel rods in dry-cask storage and determines their isotopic composition without moving or opening the cask. For more information about this project, visit http://www.inl.gov/rd100/2011/compton-dry-cask-imaging-system/

  15. High Burnup Dry Storage Cask Research and Development Project...

    Broader source: Energy.gov (indexed) [DOE]

    potentially significant impact on nuclear plant licensing and operations. While dry storage of lower burnup SNF less than 45 gigawatt days per metric ton uranium (GWD MTU)...

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

    SciTech Connect (OSTI)

    Fu-Kuang Ko; Liang, Thomas K.S.; Chung-Yu Yang [Institute of Nuclear Energy Research P.O. Box 3-3, Longtan, 32500, Taiwan (China)

    2002-07-01T23:59:59.000Z

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

  17. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    SciTech Connect (OSTI)

    none,

    2014-02-27T23:59:59.000Z

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  18. Dry Cask Storage Experience for a One-of-a-Kind Decommissioning Project

    SciTech Connect (OSTI)

    Lehnert, Robert [Energy Solutions, Spent Fuel Division, Inc: 2105 S. Bascom Ave., Suite 160, Campbell, California 95008 (United States); Trubilowicz, William [Operating Solutions of Michigan, Inc: 9039 Norton Road, Charlevoix, Michigan 49720 (United States)

    2008-01-15T23:59:59.000Z

    The Big Rock Point Restoration Project faced many unique challenges in preparation to remove all of the spent fuel from the fuel pool where it had been stored for almost thirty years to facilitate decommissioning and dismantling the entire plant. Being the first site to use a new cask system to place the fuel into dry cask storage canisters to be stored at the Independent Spent Fuel Storage Installation (ISFSI) on the Big Rock site was among the challenges. Providing the ability for cask handling operations after the spent fuel pool had been dismantled provided another challenge. The purpose of this paper is to discuss the challenges that the Big Rock team faced in completing this task on a schedule that met the goals of the Restoration Project. In conclusion, the unique features of the Big Rock plant and fuel, coupled with the goals and objectives of the Big Rock decommissioning and site restoration project posed considerable challenges that were successfully overcome by the Big Rock team. The Big Rock spent fuel was successfully moved to dry cask storage in a stand-alone ISFSI awaiting DOE to remove it from the site, and the plant structures, including the spent fuel pool, have been successfully demolished and removed from the site. The site with the exception of the ISFSI has been fully restored and was released by the NRC for unrestricted use on January 08, 2007.

  19. Managing Aging Effects on Dry Cask Storage Systems for Extended...

    Broader source: Energy.gov (indexed) [DOE]

    Electric Grid Workshop, April 19-20, 2011 Before the House Energy and Commerce Subcommittee on Energy and Power Effects of Multiple Drying Cycles on HBU PWR Cladding Alloys...

  20. Testing of the dual slab verification detector for attended measurements of the BN-350 dry storage casks

    SciTech Connect (OSTI)

    Santi, Peter A [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    The Dual Slab Verification Detector (DSVD) has been developed and built by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of {sup 3}He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. The DSVD will be used to perform measurements of the neutron flux emanating from inside the dry storage cask at several locations around each cask to establish a neutron 'fingerprint' that is sensitive to the contents of the cask. The sensitivity of the fingerprinting technique to the removal of specific amount of nuclear material from the cask is determined by the characteristics of the detector that is used to perform the measurements, the characteristics of the spent fuel being measured, and systematic uncertainties that are associated with the dry storage scenario. MCNPX calculations of the BN-350 dry storage asks and layout have shown that the neutron fingerprint verification technique using measurements from the DSVD would be sensitive to both the amount and location of material that is present within an individual cask. To confirm the performance of the neutron fingerprint technique in verifying the presence of BN-350 spent fuel in dry storage, an initial series of measurements have been performed to test the performance and characteristics of the DSVD. Results of these measurements will be presented and compared with MCNPX results.

  1. Computational Fluid Dynamics Best Practice Guidelines in the Analysis of Storage Dry Cask

    SciTech Connect (OSTI)

    Zigh, A.; Solis, J. [US Nuclear Regulatory Commission, Rockville, MD MS (United States)

    2008-07-01T23:59:59.000Z

    Computational fluid dynamics (CFD) methods are used to evaluate the thermal performance of a dry cask under long term storage conditions in accordance with NUREG-1536 [NUREG-1536, 1997]. A three-dimensional CFD model was developed and validated using data for a ventilated storage cask (VSC-17) collected by Idaho National Laboratory (INL). The developed Fluent CFD model was validated to minimize the modeling and application uncertainties. To address modeling uncertainties, the paper focused on turbulence modeling of buoyancy driven air flow. Similarly, in the application uncertainties, the pressure boundary conditions used to model the air inlet and outlet vents were investigated and validated. Different turbulence models were used to reduce the modeling uncertainty in the CFD simulation of the air flow through the annular gap between the overpack and the multi-assembly sealed basket (MSB). Among the chosen turbulence models, the validation showed that the low Reynolds k-{epsilon} and the transitional k-{omega} turbulence models predicted the measured temperatures closely. To assess the impact of pressure boundary conditions used at the air inlet and outlet channels on the application uncertainties, a sensitivity analysis of operating density was undertaken. For convergence purposes, all available commercial CFD codes include the operating density in the pressure gradient term of the momentum equation. The validation showed that the correct operating density corresponds to the density evaluated at the air inlet condition of pressure and temperature. Next, the validated CFD method was used to predict the thermal performance of an existing dry cask storage system. The evaluation uses two distinct models: a three-dimensional and an axisymmetrical representation of the cask. In the 3-D model, porous media was used to model only the volume occupied by the rodded region that is surrounded by the BWR channel box. In the axisymmetric model, porous media was used to model the entire region that encompasses the fuel assemblies as well as the gaps in between. Consequently, a larger volume is represented by porous media in the second model; hence, a higher frictional flow resistance is introduced in the momentum equations. The conservatism and the safety margins of these models were compared to assess the applicability and the realism of these two models. The three-dimensional model included fewer geometry simplifications and is recommended as it predicted less conservative fuel cladding temperature values, while still assuring the existence of adequate safety margins. (authors)

  2. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Morrow, Charles W.

    2013-01-01T23:59:59.000Z

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level - 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  3. Design review report FFTF interim storage cask

    SciTech Connect (OSTI)

    Scott, P.L.

    1995-01-03T23:59:59.000Z

    Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location.

  4. Standard review plan for reviewing safety analysis reports for dry metallic spent fuel storage casks

    SciTech Connect (OSTI)

    Not Available

    1988-01-01T23:59:59.000Z

    The Cask Standard Review Plan (CSRP) has been prepared as guidance to be used in the review of Cask Safety Analysis Reports (CSARs) for storage packages. The principal purpose of the CSRP is to assure the quality and uniformity of storage cask reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The CSRP also sets forth solutions and approaches determined to be acceptable in the past by the NRC staff in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a CSAR does not have to follow the solutions or approaches presented in the CSRP. However, applicants should recognize that the NRC staff has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches.

  5. Development of a novel ultrasonic temperature probe for long-term monitoring of dry cask storage systems

    SciTech Connect (OSTI)

    Bakhtiari, S.; Wang, K.; Elmer, T. W.; Koehl, E.; Raptis, A. C. [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL, 60439 (United States)

    2013-01-25T23:59:59.000Z

    With the recent cancellation of the Yucca Mountain repository and the limited availability of wet storage utilities for spent nuclear fuel (SNF), more attention has been directed toward dry cask storage systems (DCSSs) for long-term storage of SNF. Consequently, more stringent guidelines have been issued for the aging management of dry storage facilities that necessitate monitoring of the conditions of DCSSs. Continuous health monitoring of DCSSs based on temperature variations is one viable method for assessing the integrity of the system. In the present work, a novel ultrasonic temperature probe (UTP) is being tested for long-term online temperature monitoring of DCSSs. Its performance was evaluated and compared with type N thermocouple (NTC) and resistance temperature detector (RTD) using a small-scale dry storage canister mockup. Our preliminary results demonstrate that the UTP system developed at Argonne is able to achieve better than 0.8 Degree-Sign C accuracy, tested at temperatures of up to 400 Degree-Sign C. The temperature resolution is limited only by the sampling rate of the current system. The flexibility of the probe allows conforming to complex geometries thus making the sensor particularly suited to measurement scenarios where access is limited.

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

    SciTech Connect (OSTI)

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

    1980-08-01T23:59:59.000Z

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

  7. Initial measurements of BN-350 spent fuel in dry storage casks using the dual slab verification detonator

    SciTech Connect (OSTI)

    Santi, Peter Angelo [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Freeman, Corey R [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory

    2010-01-01T23:59:59.000Z

    The Dual Slab Verification Detector (DSVD) has been developed, built, and characterized by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of 3He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. By performing DSVD measurements at several different locations around the outer surface of the DUC, a signature 'fingerprint' can be established for each DUC based on the neutron flux emanating from inside the dry storage cask. The neutron fingerprint for each individual DUC will be dependent upon the spatial distribution of nuclear material within the cask, thus making it sensitive to the removal of a certain amount of material from the cask. An initial set of DSVD measurements have been performed on the first set of dry storage casks that have been loaded with canisters of spent fuel and moved onto the dry storage pad to both establish an initial fingerprint for these casks as well as to quantify systematic uncertainties associated with these measurements. The results from these measurements will be presented and compared with the expected results that were determined based on MCNPX simulations of the dry storage facility. The ability to safeguard spent nuclear fuel is strongly dependent on the technical capabilities of establishing and maintaining continuity of knowledge (COK) of the spent fuel as it is released from the reactor core and either reprocessed or packaged and stored at a storage facility. While the maintenance of COK is often done using continuous containment and surveillance (C/S) on the spent fuel, it is important that the measurement capabilities exist to re-establish the COK in the event of a significant gap in the continuous CIS by performing measurements that independently confirm the presence and content of Plutonium (Pu) in the spent fuel. The types of non-destructive assay (NDA) measurements that can be performed on the spent fuel are strongly dependent on the type of spent fuel that is being safeguarded as well as the location in which the spent fuel is being stored. The BN-350 Spent Fuel Disposition Project was initiated to improve the safeguards and security of the spent nuclear fuel from the BN-350 fast-breeder reactor and was developed cooperatively to meet the requirements of the International Atomic Energy Agency (IAEA) as well as the terms of the 1993 CTR and MPC&A Implementing Agreements. The unique characteristics of fuel from the BN-350 fast-breeder reactor have allowed for the development of an integrated safeguards measurement program to inventory, monitor, and if necessary, re-verify Pu content of the spent fuel throughout the lifetime of the project. This approach includes the development of a safeguards measurement program to establish and maintain the COK on the spent fuel during the repackaging and eventual relocation of the spent-fuel assemblies to a long-term storage site. As part of the safeguards measurement program, the Pu content of every spent-fuel assembly from the BN-350 reactor was directly measured and characterized while the spent-fuel assemblies were being stored in the spent-fuel pond at the BN-350 facility using the Spent Fuel Coincidence Counter (SFCC). Upon completion of the initial inventory of the Pu content of the individual spent-fuel assemblies, the assemblies were repackaged into welded steel canisters that were filled with inert argon gas and held either four or six individual spent-fuel assemblies depending on the type of assembly that was being packaged. This repackaging of the spent-fuel assemblies was performed in order to improve the stability of the spent-fuel assemblies for long-term storage and increase the proliferation resistance of the spent fuel. To maintain the capability of verifying the presence of the spent-fuel assemblies inside the welded steel canisters, measurements were performed on the canis

  8. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    SciTech Connect (OSTI)

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01T23:59:59.000Z

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

  9. Feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer based on single-element transfer cask experience

    SciTech Connect (OSTI)

    Schmoker, D.S.; Bowser, R.C.

    1993-12-31T23:59:59.000Z

    Spent fuel transportation casks and canister-based storage systems are generally loaded underwater in a nuclear plant`s spent fuel pool/cask loading pit. Several reasons exist for exploring the feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer. These include: the accommodation of plants which do not have sufficient crane capacity to handle large 90 tonne (100 ton) storage canisters or shipping casks, and construction of an MRS without the need for extensive hot cell facilities. In the case of DOE`s ``Multi-Purpose Canister`` (MPC) scenario, use of such a transfer system would allow all plants with adequate transport routes to use large canisters at-reactor, and those without adequate transport routes to use the MRS for loading of large canisters without the need for hot cell facilities. The dry transfer option would not only allow the use of large canisters for all fuel, but would assist DOE in meeting MRS deadlines since licensing and construction of hot-cell facilities significantly affect schedule. This paper reviews the regulatory issues and technical design considerations for a single-element dry transfer system. Also summarized are lessons learned from the TMI-2 fuel transfer system which are directly applicable to the design, testing, startup, and use of a future dry cask-to-cask or pool-to-cask transfer system.

  10. Managing aging effects on dry cask storage systems for extended long-term storage and transportation of used fuel - rev. 0

    SciTech Connect (OSTI)

    Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y. (Decision and Information Sciences); ( EVS); ( NE)

    2012-07-06T23:59:59.000Z

    The cancellation of the Yucca Mountain repository program in the United States raises the prospect of extended long-term storage (i.e., >120 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Application for ISFSI license renewals must include the following: (1) Time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation; and (2) a description of the aging management program (AMP) for management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must also include design bases information as documented in the most recent updated final safety analysis report as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference provided that those references are clear and specific. The NRC has recently issued the Standard Review Plan (SRP) for renewal of used-fuel dry cask storage system (DCSS) licenses and Certificates of Compliance (CoCs), NUREG-1927, under which NRC may renew a specific license or a CoC for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and DCSS that address aging effects that could affect the safe storage of the used fuel. The information contained in the license and CoC renewal applications will require NRC review to verify that the aging effects on the SSCs in DCSSs/ ISFSIs are adequately managed for the period of extended operation. To date, all of the ISFSIs located across the United States with more than 1,500 dry casks loaded with used fuel have initial license terms of 20 years; three ISFSIs (Surry, H.B. Robinson and Oconee) have received their renewed licenses for 20 years, and two other ISFSIs (Calvert Cliffs and Prairie Island) have applied for license renewal for 40 years. This report examines issues related to managing aging effects on the SSCs in DCSSs/ISFSIs for extended long-term storage and transportation of used fuels, following an approach similar to that of the Generic Aging Lessons Learned (GALL) report, NUREG-1801, for the aging management and license renewal of nuclear power plants. The report contains five chapters and an appendix on quality assurance for aging management programs for used-fuel dry storage systems. Chapter I of the report provides an overview of the ISFSI license renewal process based on 10 CFR 72 and the guidance provided in NUREG-1927. Chapter II contains definitions and terms for structures and components in DCSSs, materials, environments, aging effects, and aging mechanisms. Chapter III and Chapter IV contain generic TLAAs and AMPs, respectively, that have been developed for managing aging effects on the SSCs important to safety in the dry cask storage system designs described in Chapter V. The summary descriptions and tabulations of evaluations of AMPs and TLAAs for the SSCs that are important to safety in Chapter V include DCSS designs (i.e., NUHOMS{reg_sign}, HI-STORM 100, Transnuclear (TN) metal cask, NAC International S/T storage cask, ventilated storage cask (VSC-24), and the Westinghouse MC-10 metal dry storage cask) that have been and continue to be used by utilities across the country for dry storage of used fuel to date. The goal of this report is to help establish the technical

  11. Dose Rates for Various Loading Patterns of Spent Fuel Assemblies in a Dry Cask

    SciTech Connect (OSTI)

    Jenquin, Urban P. (BATTELLE (PACIFIC NW LAB))

    2001-01-01T23:59:59.000Z

    Shielding calculations were performed to assess the impact of loading various combinations of spent fuel on dose rates and fuel temperature in a dry storage cask.

  12. Managing Aging Effects on Dry Cask Storage Systems for Extended Long Term

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(FactDepartment ofLetterEconomyDr.Energy University ofOverviewManagementStorage

  13. Source storage and transfer cask: Users Guide

    SciTech Connect (OSTI)

    Eccleston, G.W.; Speir, L.G.; Garcia, D.C.

    1985-04-01T23:59:59.000Z

    The storage and shield cask for the dual californium source is designed to shield and transport up to 3.7 mg (2 Ci) of /sup 252/Cf. the cask meets Department of Transportation (DOT) license requirements for Type A materials (DOT-7A). The cask is designed to transfer sources to and from the Flourinel and Fuel Storage (FAST) facility delayed-neutron interrogator. Californium sources placed in the cask must be encapsulated in the SR-CF-100 package and attached to Teleflex cables. The cask contains two source locations. Each location contains a gear box that allows a Teleflex cable to be remotely moved by a hand crank into and out of the cask. This transfer procedure permits sources to be easily removed and inserted into the delayed-neutron interrogator and reduces personnel radiation exposure during transfer. The radiation dose rate with the maximum allowable quantity of californium (3.7 mg) in the cask is 30 mR/h at the surface and less than 2 mR/h 1 m from the cask surface. This manual contains information about the cask, californium sources, describes the method to ship the cask, and how to insert and remove sources from the cask. 28 figs.

  14. IMPACT ANALYSIS OF SPENT FUEL DRY CASKS UNDER ACCIDENTAL DROP SCENARIOS.

    SciTech Connect (OSTI)

    BRAVERMAN,J.I.; MORANTE,R.J.; XU,J.; HOFMAYER,C.H.; SHAUKAT,S.K.

    2003-03-17T23:59:59.000Z

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister (MPC) that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel.

  15. IMPACT ANALYSIS OF SPENT FUEL DRY CASKS UNDER ACCIDENTAL DROP SCENARIOS.

    SciTech Connect (OSTI)

    BRAVERMAN,J.I.; MORANTE,R.J.; XU,J.; HOFMAYER,C.H.; SHAUKAT,S.K.

    2003-08-17T23:59:59.000Z

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister OLIIpC that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel.

  16. A Review of NDE Methods for Detecting and Monitoring of Atmospheric SCC in Dry Cask Storage Canisters for Used Nuclear Fuel

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Hanson, Brady D.; Sorenson, Ken B.

    2013-04-01T23:59:59.000Z

    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.

  17. RADIATION ANALYSIS OF A SPENT-FUEL STORAGE CASK

    E-Print Network [OSTI]

    Shultis, J. Kenneth

    RADIATION ANALYSIS OF A SPENT-FUEL STORAGE CASK by J.K. Shultis Department of Mechanical;Radiation Analysis of a Spect-Fuel Storage Cask by J.K.Shultis Dept. Mechanical and Nuclear Engineering a single Transnuclear spent-fuel storage cask holding 68 design-basis fuel assemblies (a TN-68 cask

  18. Spent fuel integrity during dry storage

    SciTech Connect (OSTI)

    McKinnon, M.A.

    1995-07-01T23:59:59.000Z

    Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at the Idaho National Engineering Laboratory (INEL) offers significant opportunities for confirmation of the benign nature of long-term dry storage. The cask performance tests conducted at INEL included visual observation and ultrasonic examination of the condition of cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of the fuel; and a qualitative determination of the effect of dry storage and fuel consolidation on fission gas release from the spent fuel rods. A variety of cover gases and cask orientations were used during the cask performance tests. Cover gases included vacuum, nitrogen, and helium. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the conclusion of each performance test, periodic gas sampling was conducted on each cask as part of a surveillance and monitoring activity. Continued surveillance and monitoring activities are being conducted for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are reported in this paper.

  19. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01T23:59:59.000Z

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  20. Study on concrete cask storage of spent fuel in Japan

    SciTech Connect (OSTI)

    Itoh, C. [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Nuclear Fuel Cycle Dept.; Onodera, A.; Yamada, N. [Hitachi Zosen Corp., Tokyo (Japan). Nuclear Div.

    1993-12-31T23:59:59.000Z

    The present report describes the status of the first year`s work of a five-year-long study on concrete cask storage of spent fuel in Japan. Firstly, the proposed study program is elaborated to clarify the position of the present work. Then, the results of the study which have been obtained so far are described and the technical issues are addressed to make the concrete cask storage viable in Japan.

  1. Horizontal modular dry irradiated fuel storage system

    DOE Patents [OSTI]

    Fischer, Larry E. (Los Gatos, CA); McInnes, Ian D. (San Jose, CA); Massey, John V. (San Jose, CA)

    1988-01-01T23:59:59.000Z

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

  2. Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

    SciTech Connect (OSTI)

    Krementz, Dan; Rose, David; Dunsmuir, Mike

    2014-02-06T23:59:59.000Z

    The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit #2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24” diameter.[1, 2] This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18” in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit #2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24” in diameter and ~11 feet long from a dry transfer cask to the basin. The 18” and 24” applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit #2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with different diameters and lengths would likely be on the same order of magnitude as the Basin Modifications project. The cost of a DTS capability is affected by the number of design variations of different vendor transport and dry transfer casks to be considered for design input. Some costs would be incurred for each vendor DTS to be handled. For example, separate analyses would be needed for each dry transfer cask type such as criticality, shielding, dropping a dry transfer cask and basket, handling and auxiliary equipment, procedures, operator training, readiness assessments, and operational readiness reviews. A DTS handling capability in L-Area could serve as a backup to the Shielded Transfer System (STS) for unloading long casks and could support potential future missions such as the Idaho National Laboratory (INL) Exchange or transferring UNF from wet to dry storage.

  3. PRELIMINARY REPORT: EFFECTS OF IRRADIATION AND THERMAL EXPOSURE ON ELASTOMERIC SEALS FOR CASK TRANSPORTATION AND STORAGE

    SciTech Connect (OSTI)

    Verst, C.; Skidmore, E.; Daugherty, W.

    2014-05-30T23:59:59.000Z

    A testing and analysis approach to predict the sealing behavior of elastomeric seal materials in dry storage casks and evaluate their ability to maintain a seal under thermal and radiation exposure conditions of extended storage and beyond was developed, and initial tests have been conducted. The initial tests evaluate the aging response of EPDM elastomer O-ring seals. The thermal and radiation exposure conditions of the CASTOR® V/21 casks were selected for testing as this cask design is of interest due to its widespread use, and close proximity of the seals to the fuel compared to other cask designs leading to a relatively high temperature and dose under storage conditions. A novel test fixture was developed to enable compression stress relaxation measurements for the seal material at the thermal and radiation exposure conditions. A loss of compression stress of 90% is suggested as the threshold at which sealing ability of an elastomeric seal would be lost. Previous studies have shown this value to be conservative to actual leakage failure for most aging conditions. These initial results indicate that the seal would be expected to retain sealing ability throughout extended storage at the cask design conditions, though longer exposure times are needed to validate this assumption. The high constant dose rate used in the testing is not prototypic of the decreasingly low dose rate that would occur under extended storage. The primary degradation mechanism of oxidation of polymeric compounds is highly dependent on temperature and time of exposure, and with radiation expected to exacerbate the oxidation.

  4. Standard practice for qualification and acceptance of boron based metallic neutron absorbers for nuclear criticality control for dry cask storage systems and transportation packaging

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2007-01-01T23:59:59.000Z

    1.1 This practice provides procedures for qualification and acceptance of neutron absorber materials used to provide criticality control by absorbing thermal neutrons in systems designed for nuclear fuel storage, transportation, or both. 1.2 This practice is limited to neutron absorber materials consisting of metal alloys, metal matrix composites (MMCs), and cermets, clad or unclad, containing the neutron absorber boron-10 (10B). 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  5. Research on Spent Fuel Storage and Transportation in CRIEPI (Part 2 Concrete Cask Storage)

    SciTech Connect (OSTI)

    Koji Shirai; Jyunichi Tani; Taku Arai; Masumi Watatu; Hirofumi Takeda; Toshiari Saegusa; Philip L. Winston

    2008-10-01T23:59:59.000Z

    Concrete cask storage has been implemented in the world. At a later stage of storage period, the containment of the canister may deteriorate due to stress corrosion cracking phenomena in a salty air environment. High resistant stainless steels against SCC have been tested as compared with normal stainless steel. Taking account of the limited time-length of environment with certain level of humidity and temperature range, the high resistant stainless steels will survive from SCC damage. In addition, the adhesion of salt from salty environment on the canister surface will be further limited with respect to the canister temperature and angle of the canister surface against the salty air flow in the concrete cask. Optional countermeasure against SCC with respect to salty air environment has been studied. Devices consisting of various water trays to trap salty particles from the salty air were designed to be attached at the air inlet for natural cooling of the cask storage building. Efficiency for trapping salty particles was evaluated. Inspection of canister surface was carried out using an optical camera inserted from the air outlet through the annulus of a concrete cask that has stored real spent fuel for more than 15 years. The camera image revealed no gross degradation on the surface of the canister. Seismic response of a full-scale concrete cask with simulated spent fuel assemblies has been demonstrated. The cask did not tip over, but laterally moved by the earthquake motion. Stress generated on the surface of the spent fuel assemblies during the earthquake motion were within the elastic region.

  6. BWR - Spent Fuel Transport and Storage with the TN{sup TM}9/4 and TN{sup TM}24BH Casks

    SciTech Connect (OSTI)

    Wattez, L. [COGEMA LOGISTICS - AREVA Group (France); Marguerat, Y. [BKW FMB Energy Ltd (Switzerland); Hoesli, C. [ZWILAG Zwischenlager Wuerenlingen AG (Switzerland)

    2006-07-01T23:59:59.000Z

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN{sup TM}24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN{sup TM}9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN{sup TM}9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN{sup TM}24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN{sup TM}9/4 round trips are performed, and one TN{sup TM}24BH is loaded. 5 additional TN{sup TM}24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN{sup TM}24BH high capacity dual purpose cask and the TN{sup TM}9/4 transport cask characteristics and capabilities will then be detailed. (authors)

  7. Measurement of Atmospheric Sea Salt Concentration in the Dry Storage Facility of the Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Masumi Wataru; Hisashi Kato; Satoshi Kudo; Naoko Oshima; Koji Wada [Central Research Institute of Electric Power Industry - CRIEPI (Japan); Hirofumi Narutaki [Electric Power Engineering Systems Co. Ltd. (Japan)

    2006-07-01T23:59:59.000Z

    Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility. Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack. (authors)

  8. CASTOR cask with high loading capacity for transport and storage of VVER 440 spent fuel

    SciTech Connect (OSTI)

    Diersch, R.; Methling, D.; Milde, G. [Gesellschaft fuer Nuklear-Behaelter mbH Essen (Germany)

    1993-12-31T23:59:59.000Z

    GNB has developed a CASTOR transport and storage cask with a capacity of 84 spent fuel assemblies from reactors of the type VVER 440. The safety analyses are performed with the help of modern, benchmarked calculation programs. The results show that the cask design is able to fulfill both the Type B test conditions on basis of IAEA Regulations-1985 edition and the requirements for interim storage sites in Germany.

  9. Extending dry storage of spent LWR fuel for 100 years.

    SciTech Connect (OSTI)

    Einziger, R. E.

    1998-12-16T23:59:59.000Z

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and excessive creep and mechanical property changes. Postulated accident scenarios would be the same for 20-year or 100-year storage, because they are mostly governed by operational or outside events, and not by the cask or fuel. Analyses of accident scenarios during extended dry storage could be impacted by fuel and cask changes that would result from the extended period of storage. Overall, the results of this work indicate that, based on fuel behavior, spent fuel at burnups below {approximately}45 GWd/MTU can be dry stored for 100 years. Long-term storage of higher burnup fuel or fuels with newer cladding will require the determination of temperature limits based on evaluation of stress-driven degradation mechanisms of the cladding.

  10. Drying Rough Rice in Storage.

    E-Print Network [OSTI]

    Sorenson, J. W. Jr.; Crane, L. E.

    1960-01-01T23:59:59.000Z

    Drying. Rough Rice in Storage Ih AGRf""' TURP YPERIMENT STAT10 I. TEXAS SUMMARY Research was conducted at the Rice-Pasture Experiment Station near Beaumont during 7 crop years (1952-53 through 1958-59) to determine the engineering problems... and the practicability of dry- ing rough rice in storage in Texas. Drying rice in storage means drying rice in the same bin in which it is to be stored. Rough rice, with initial moisture contents of 15.0 to 23.0 percent, was dried at depths of 4 to 10 feet...

  11. Use of transportable storage casks in the nuclear waste management system: Appendices

    SciTech Connect (OSTI)

    Not Available

    1987-12-01T23:59:59.000Z

    A study was performed to determine the viability of the use of transportable storage casks (TSCs), and other metal casks that are designed primarily for storage but which might be used to ship their stored contents to DOE on a one-time use basis (referred to in this study as storage only casks, or SOCs), in the combined utility/DOE spent fuel management system. The viability of the use of TSCs and SOCs was assessed in terms of the costs and savings involved in their use, the sensitivity of these costs and savings to changes in the capacity and cost of fabrication of the casks, the impacts of variation in cask design features on cost and radiation exposure of personnel, and their prospective use in connection with the transport of defense high level wastes. Estimates were developed of the costs of acquiring and handling of TSCs and SOCs at reactor sites. For comparison purposes, similar costs were developed for the use of concrete storage casks at reactor sites. Estimates of the savings involved to the DOE system as a result of receiving spent fuel in TSCs or SOCs were separately developed. These costs are developed and presented in Volume 2, Appendices A through J.

  12. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect (OSTI)

    Romanato, L.S. [Centro Tecnologico da Marinha em S.Paulo, Brazilian Navy Technological Center, Sao Paulo (Brazil)

    2008-07-01T23:59:59.000Z

    This paper aims to present the advantages of dry cask storage in comparison with the wet storage (cooling water pools) for SNF. When the nuclear fuel is removed from the core reactor, it is moved to a storage unit and it wait for a final destination. Generally, the spent nuclear fuel (SNF) remains inside water pools within the reactors facility for the radioactive activity decay. After some period of time in pools, SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing facilities, or still, wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet facilities, depending on the method adopted by the nuclear power plant or other plans of the country. Interim storage, up to 20 years ago, was exclusively wet and if the nuclear facility had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry facilities. This kind of storage does not need complex radiation monitoring and it is safer then wet one. Casks, either concrete or metallic, are safer, especially on occurrence of earthquakes, like that occurred at Kashiwazaki-Kariwa nuclear power plant, in Japan on July 16, 2007. (authors)

  13. Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doherty, A.L.

    1997-06-01T23:59:59.000Z

    This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.

  14. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    SciTech Connect (OSTI)

    Bailey, W.J.

    1990-02-01T23:59:59.000Z

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  15. CHARACTERISTICS OF NEXT-GENERATION SPENT NUCLEAR FUEL (SNF) TRANSPORT AND STORAGE CASKS

    SciTech Connect (OSTI)

    Haire, M.J.; Forsberg, C.W.; Matveev, V.Z.; Shapovalov, V.I.

    2004-10-03T23:59:59.000Z

    The design of spent nuclear fuel (SNF) casks used in the present SNF disposition systems has evolved from early concepts about the nuclear fuel cycle. The reality today is much different from that envisioned by early nuclear scientists. Most SNF is placed in pool storage, awaiting reprocessing (as in Russia) or disposal at a geologic SNF repository (as in the United States). Very little transport of SNF occurs. This paper examines the requirements for SNF casks from today's perspective and attempts to answer this question: What type of SNF cask would be produced if we were to start over and design SNF casks based on today's requirements? The characteristics for a next-generation SNF cask system are examined and are found to be essentially the same in Russia and the United States. It appears that the new depleted uranium dioxide (DUO2)-steel cermet material will enable these requirements to be met. Depleted uranium (DU) is uranium in which a portion of the 235U isotope has been removed during a uranium enrichment process. The DUO2-steel cermet material is described. The United States and Russia are cooperating toward the development of a next-generation, dual-purpose, storage and transport SNF system.

  16. Structural Sensitivity of Dry Storage Canisters

    SciTech Connect (OSTI)

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

    2013-09-27T23:59:59.000Z

    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 raising the drop height or changing the impact angle. The seismic load case was derived from the August 23, 2011 earthquake that affected the North Anna power station. The source of the data was a monitoring station near Charlottesville, Virginia, so the ground motion is not an exact match. Stresses on the containment boundary were so low, even from a fatigue standpoint, that the seismic load case is generally not a concern. Based on this study, it is recommended that high strain rate testing of welded stainless steel test samples be pursued to define the currently unknown material behavior. Additional modeling is recommended to evaluate specific dry storage cask system designs subjected to tip-over loads using a high level of model detail. Additional modeling of the canister interior components (basket, fuel assemblies, etc.) is also recommended, to evaluate the feasibility of fuel retrievability after a tip-over incident. Finally, additional modeling to determine how much degradation a system could undergo and still maintain the integrity of the confinement barrier should be performed.

  17. Nondestructive Evaluation of the VSC-17 Cask

    SciTech Connect (OSTI)

    Sheryl Morton; Al Carlson; Cecilia Hoffman; James Rivera; Phil Winston; Koji Shirai; Shin Takahashi; Masaharo Tanaka

    2006-01-01T23:59:59.000Z

    In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC 17) spent nuclear fuel storage cask, originally located at the INL Test Area North, as a candidate to study cask performance because it had been used to store fuel as part of a dry cask storage demonstration project for over 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. The INL team met with the CRIEPI representatives in December of 2004 to discuss the next steps. As a result of that meeting, CRIEPI requested that in the summer 2005 INL perform additional surveys on the VSC 17 cask with participation of CRIEPI scientists. This document summarizes the evaluation methods used on the VSC 17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.

  18. Demonstrating the Safety of Long-Term Dry Storage - 13468

    SciTech Connect (OSTI)

    McCullum, Rod [Nuclear Energy Institute, 1201 F St. NW, Washington, DC, 20004 (United States)] [Nuclear Energy Institute, 1201 F St. NW, Washington, DC, 20004 (United States); Brookmire, Tom [Dominion Energy, 5000 Dominion Boulevard Glen Allen, VA 23060 (United States)] [Dominion Energy, 5000 Dominion Boulevard Glen Allen, VA 23060 (United States); Kessler, John [Electric Power Research Institute, 1300 West W.T. Harris Boulevard, Charlotte, NC 28262 (United States)] [Electric Power Research Institute, 1300 West W.T. Harris Boulevard, Charlotte, NC 28262 (United States); Leblang, Suzanne [Entergy, 1340 Echelon Parkway, Jackson, MS 39211 (United States)] [Entergy, 1340 Echelon Parkway, Jackson, MS 39211 (United States); Levin, Adam [Exelon, 4300 Winfield Road, Warrenville, IL 60555 (United States)] [Exelon, 4300 Winfield Road, Warrenville, IL 60555 (United States); Martin, Zita [Tennessee Valley Authority, 1101 Market Street, Chattanooga, TN 37402 (United States)] [Tennessee Valley Authority, 1101 Market Street, Chattanooga, TN 37402 (United States); Nesbit, Steve [Duke Energy, 550 South Tryon Street, Charlotte, NC 28202 (United States)] [Duke Energy, 550 South Tryon Street, Charlotte, NC 28202 (United States); Nichol, Marc [Nuclear Energy Institute, 1201 F St. NW Washington DC, 2004 (United States)] [Nuclear Energy Institute, 1201 F St. NW Washington DC, 2004 (United States); Pickens, Terry [Xcel Energy, 414 Nicollet Mall, Minneapolis, MN 55401 (United States)] [Xcel Energy, 414 Nicollet Mall, Minneapolis, MN 55401 (United States)

    2013-07-01T23:59:59.000Z

    Commercial nuclear plants in the United States were originally designed with the expectation that used nuclear fuel would be moved directly from the reactor pools and transported off site for either reprocessing or direct geologic disposal. However, Federal programs intended to meet this expectation were never able to develop the capability to remove used fuel from reactor sites - and these programs remain stalled to this day. Therefore, in the 1980's, with reactor pools reaching capacity limits, industry began developing dry cask storage technology to provide for additional on-site storage. Use of this technology has expanded significantly since then, and has today become a standard part of plant operations at most US nuclear sites. As this expansion was underway, Federal programs remained stalled, and it became evident that dry cask systems would be in use longer than originally envisioned. In response to this challenge, a strong technical basis supporting the long term dry storage safety has been developed. However, this is not a static situation. The technical basis must be able to address future challenges. Industry is responding to one such challenge - the increasing prevalence of high burnup (HBU) used fuel and the need to provide long term storage assurance for these fuels equivalent to that which has existed for lower burnup fuels over the past 25 years. This response includes a confirmatory demonstration program designed to address the aging characteristics of HBU fuel and set a precedent for a learning approach to aging management that will have broad applicability across the used fuel storage landscape. (authors)

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

    SciTech Connect (OSTI)

    Botsch, W.; Smalian, S.; Hinterding, P. [TUV NORD Nuclear c/o TUV NORD EnSys Hannover GmbH and Co.KG, Dept. Radiation Protection and Waste Disposal, Am TueV 1, 30519 Hannover (Germany)] [TUV NORD Nuclear c/o TUV NORD EnSys Hannover GmbH and Co.KG, Dept. Radiation Protection and Waste Disposal, Am TueV 1, 30519 Hannover (Germany); Voelzke, H.; Wolff, D.; Kasparek, E. [BAM Federal Institute for Materials Research and Testing Division 3.4 Safety of Storage Containers Unter den Eichen 44-46, 12203 Berlin (Germany)] [BAM Federal Institute for Materials Research and Testing Division 3.4 Safety of Storage Containers Unter den Eichen 44-46, 12203 Berlin (Germany)

    2013-07-01T23:59:59.000Z

    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 casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

  20. Use of depleted uranium metal as cask shielding in high-level waste storage, transport, and disposal systems

    SciTech Connect (OSTI)

    Yoshimura, H.R.; Ludwigsen, J.S.; McAllaster, M.E. [and others

    1996-09-01T23:59:59.000Z

    The US DOE has amassed over 555,000 metric tons of depleted uranium from its uranium enrichment operations. Rather than dispose of this depleted uranium as waste, this study explores a beneficial use of depleted uranium as metal shielding in casks designed to contain canisters of vitrified high-level waste. Two high-level waste storage, transport, and disposal shielded cask systems are analyzed. The first system employs a shielded storage and disposal cask having a separate reusable transportation overpack. The second system employs a shielded combined storage, transport, and disposal cask. Conceptual cask designs that hold 1, 3, 4 and 7 high-level waste canisters are described for both systems. In all cases, cask design feasibility was established and analyses indicate that these casks meet applicable thermal, structural, shielding, and contact-handled requirements. Depleted uranium metal casting, fabrication, environmental, and radiation compatibility considerations are discussed and found to pose no serious implementation problems. About one-fourth of the depleted uranium inventory would be used to produce the casks required to store and dispose of the nearly 15,400 high-level waste canisters that would be produced. This study estimates the total-system cost for the preferred 7-canister storage and disposal configuration having a separate transportation overpack would be $6.3 billion. When credits are taken for depleted uranium disposal cost, a cost that would be avoided if depleted uranium were used as cask shielding material rather than disposed of as waste, total system net costs are between $3.8 billion and $5.5 billion.

  1. Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

    SciTech Connect (OSTI)

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)] [Moore Nuclear Energy, LLC (United States)

    2013-07-01T23:59:59.000Z

    Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

  2. NAC-1 cask dose rate calculations for LWR spent fuel

    SciTech Connect (OSTI)

    CARLSON, A.B.

    1999-02-24T23:59:59.000Z

    A Nuclear Assurance Corporation nuclear fuel transport cask, NAC-1, is being considered as a transport and storage option for spent nuclear fuel located in the B-Cell of the 324 Building. The loaded casks will be shipped to the 200 East Area Interim Storage Area for dry interim storage. Several calculations were performed to assess the photon and neutron dose rates. This report describes the analytical methods, models, and results of this investigation.

  3. REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)

    SciTech Connect (OSTI)

    CHASTAIN, S.A.

    2005-10-24T23:59:59.000Z

    Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The report also identified additional components and actions in Section 3.0 and Table 3 that require further evaluation. The purpose of this report is to evaluate another portion of the remaining inventory (i.e., delayed neutron signal fuel, blanket assemblies, highly enriched assemblies, newly loaded Ident-69 pin containers, and returned fuel) to ensure it can be safely off loaded to the FFTF spent fuel storage system.

  4. The Evolution of Dry Spent Fuel Storage in the United States

    SciTech Connect (OSTI)

    McGough, M.S. [Duratek Inc., 695 Bamesley Lane, Alpharetta, GA 30022 (United States); Bland, D.W. [TriVis, Inc., 1001 Yeager Parkway, Pelham, AL 35124 (United States)

    2006-07-01T23:59:59.000Z

    This paper reviews the evolution of Dry Spent Fuel storage technology and application in the United States. Dating back to the legislation signed by Jimmy Carter on April 7, 1977, to outlaw spent fuel reprocessing, the nations spent fuel pools are gradually becoming filled to capacity. This has necessitated the development of new technologies to store spent fuel in dry casks, predominantly at nuclear power plant sites, awaiting the availability of the federal repository at Yucca Mountain. Site-specific conditions and changes in types of fuel being discharged from reactors have driven a constant evolution of technologies to support this critical need. This paper provides an overview of those changes, which have influenced the evolution of dry storage technology. Focus is provided more towards current technology and cask loading practices, as opposed to those technologies, which are no longer in heavy use. Detailed pictorial material is presented showing the loading sequences of various systems in current use. This paper provides a critical primer on Dry Spent Fuel Storage technology. It provides anyone who is new to dry storage, or who is contemplating initiating dry storage at a nuclear plant site, with useful background and history upon which to build programmatic decisions. (authors)

  5. Examination of Spent Pressurized Water Reactor Fuel Rods After 15 Years in Dry Storage

    SciTech Connect (OSTI)

    Einziger, Robert E. [Argonne National Laboratory (United States); Tsai Hanchung [Argonne National Laboratory (United States); Billone, Michael C. [Argonne National Laboratory (United States); Hilton, Bruce A. [Argonne National Laboratory-West (United States)

    2003-11-15T23:59:59.000Z

    For [approximately equal to]15 yr Dominion Generation's Surry Nuclear Station 15 x 15 Westinghouse pressurized water reactor (PWR) fuel was stored in a dry inert-atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory at peak cladding temperatures that decreased from {approx}350 to 150 deg. C. Before storage, the loaded cask was subjected to thermal-benchmark tests, during which time the peak temperatures were greater than 400 deg. C. The cask was opened to examine the fuel rods for degradation and to determine if they were suitable for extended storage. No fuel rod breaches and no visible degradation or crud/oxide spallation from the fuel rod surface were observed. The results from profilometry, gas release measurements, metallographic examinations, microhardness determination, and cladding hydrogen behavior are reported in this paper.It appears that little or no fission gas was released from the fuel pellets during either the thermal-benchmark tests or the long-term storage. In the central region of the fuel column, where the axial temperature gradient in storage is small, the measured hydrogen content in the cladding is consistent with the thickness of the oxide layer. At {approx}1 m above the fuel midplane, where a steep temperature gradient existed in the cask, less hydrogen is present than would be expected from the oxide thickness that developed in-reactor. Migration of hydrogen during dry storage probably occurred and may signal a higher-than-expected concentration at the cooler ends of the rod. The volume of hydrides varies azimuthally around the cladding, and at some elevations, the hydrides appear to have segregated somewhat to the inner and outer cladding surfaces. It is, however, impossible to determine if this segregation occurred in-reactor or during transportation, thermal-benchmark tests, or the dry storage period. The hydrides retained the circumferential orientation typical of prestorage PWR fuel rods. Little or no cladding creep occurred during thermal-benchmark testing and dry storage. It is anticipated that the creep would not increase significantly during additional storage because of the lower temperature after 15 yr, continual decrease in temperature from the reduction in decay heat, and concurrent reductions in internal rod pressure and stress. This paper describes the results of the characterization of the fuel and intact cladding, as well as the implications of these results for long-term (i.e., beyond 20 yr) dry-cask storage.

  6. Conceptual design report for the ICPP spent nuclear fuel dry storage project

    SciTech Connect (OSTI)

    NONE

    1996-07-01T23:59:59.000Z

    The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

  7. Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

    SciTech Connect (OSTI)

    Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa; Koji Shirai; Akihiro Sasahara; Takatoshi Hattori

    2006-04-01T23:59:59.000Z

    In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstration project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.

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

    SciTech Connect (OSTI)

    Rasmussen, D.E.

    1982-12-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2013-08-01T23:59:59.000Z

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

  10. Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai [Ishikawajima-Harima Heavy Industries Company Ltd., 1 Shin-Nakaharacho, Isogoku, Yokohama 235-8501 (Japan)

    2002-07-01T23:59:59.000Z

    Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis was to estimate the concrete temperature increase in the case a canister contacts with guide rails in normal storage. It has a possibility that a canister contacts with guide rails during storage period after concrete cask is upended from transfer operation. The temperature increase due to this contact was calculated 5 deg. C at small local area. This result implies that the affect of the contact is very small. This paper addresses that the storage cask concrete is kept its integrity in transfer operation period and normal storage period. (authors)

  11. Inspection and Gamma-Ray Dose Rate Measurements of the Annulus of the VSC-17 Concrete Spent Nuclear Fuel Storage Cask

    SciTech Connect (OSTI)

    P. L. Winston

    2007-09-01T23:59:59.000Z

    The air cooling annulus of the Ventilated Storage Cask (VSC)-17 spent fuel storage cask was inspected using a Toshiba 7 mm (1/4”) CCD video camera. The dose rates observed in the annular space were measured to provide a reference for the activity to which the camera(s) being tested were being exposed. No gross degradation, pitting, or general corrosion was observed.

  12. Critical Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion Controlled Cavity Growth

    SciTech Connect (OSTI)

    Hayes, T.A.; Rosen, R.S.; Kassner, M.E.

    1999-12-01T23:59:59.000Z

    Interim dry storage of spent nuclear fuel (SNF) rods is of critical concern because a shortage of existing SNF wet storage capacity combined with delays in the availability of a permanent disposal repository has led to an increasing number of SNF rods being placed into interim dry storage. Safe interim dry storage must be maintained for a minimum of twenty years according to the Standard Review Plan for Dry Cask Storage Systems [1] and the Code of Federal Regulations, 10 CFR Part 72 [2]. Interim dry storage licensees must meet certain safety conditions when storing SNF rods to ensure that there is a ''very low probability (e.g. 0.5%) of cladding breach during long-term storage'' [1]. Commercial SNF typically consists of uranium oxide pellets surrounded by a thin cladding. The cladding is usually an {alpha}-zirconium based alloy know as ''Zircaloy''. In dry storage, the SNF rods are confined in one of several types of cask systems approved by the Nuclear Regulatory Commission (NRC). ''The cask system must be designed to prevent degradation of fuel cladding that results in a type of cladding breach, such as axial-splits or ductile fracture, where irradiated UO{sub 2} particles may be released. In addition, the fuel cladding should not degrade to the point where more than one percent of the fuel rods suffer pinhole or hairline crack type failure under normal storage conditions [1].'' The NRC has approved two models [3,4] for use by proposed dry storage licensees to determine the maximum initial temperature limit for nuclear fuel rods in dry storage that supposedly meet the above criteria and yield consistent temperature limits. Though these two models are based on the same fundamental failure theory, different assumptions have been made including the choice of values for material constants in the failure equation. This report will examine and compare the similarities and inconsistencies of these two models. It will illustrate some of the shortcomings of the current models and suggest modifications as well as some experiments that should be started in the near future. This report will also discuss changes in the current NRC standards with regard to the adoption of a strain-based model to be used to determine maximum allowable temperatures of the SNF.

  13. A methodology to quantify the release of spent nuclear fuel from dry casks during security-related scenarios.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Luna, Robert Earl

    2013-11-01T23:59:59.000Z

    Assessing the risk to the public and the environment from a release of radioactive material produced by accidental or purposeful forces/environments is an important aspect of the regulatory process in many facets of the nuclear industry. In particular, the transport and storage of radioactive materials is of particular concern to the public, especially with regard to potential sabotage acts that might be undertaken by terror groups to cause injuries, panic, and/or economic consequences to a nation. For many such postulated attacks, no breach in the robust cask or storage module containment is expected to occur. However, there exists evidence that some hypothetical attack modes can penetrate and cause a release of radioactive material. This report is intended as an unclassified overview of the methodology for release estimation as well as a guide to useful resource data from unclassified sources and relevant analysis methods for the estimation process.

  14. Evaluation of Cask Drop Criticality Issues at K Basin

    SciTech Connect (OSTI)

    GOLDMANN, L.H.

    2000-01-24T23:59:59.000Z

    An analysis of ability of Multi-canister Overpack (MCO) to withstand drops at K Basin without exceeding the criticality design requirements. Report concludes the MCO will function acceptably. The spent fuel currently residing in the 105 KE and 105 KW storage basins will be placed in fuel storage baskets which will be loaded into the MCO cask assembly. During the basket loading operations the MCO cask assembly will be positioned near the bottom of the south load out pit (SLOP). The loaded MCO cask will be lifted from the SLOP transferred to the transport trailer and delivered to the Cold Vacuum Drying Facility (CVDF). In the wet condition there is a potential for criticality problems if significant changes in the designed fuel configurations occur. The purpose of this report is to address structural issues associated with criticality design features for MCO cask drop accidents in the 105 KE and 105 KW facilities.

  15. Extended Dry Storage of Used Nuclear Fuel: Technical Issues: A USA Perspective

    SciTech Connect (OSTI)

    McConnell, Paul; Hanson, Brady D.; Lee, Moo; Sorenson, Ken B.

    2011-10-28T23:59:59.000Z

    Used nuclear fuel will likely be stored dry for extended periods of time in the USA. Until a final disposition pathway is chosen, the storage periods will almost definitely be longer than were originally intended. The ability of the important-to-safety structures, systems and components (SSC's) continue to meet storage and transport safety functions over extended times must be determined. It must be assured that there is no significant degradation of the fuel or dry cask storage systems. Also, it is projected that the maximum discharge burnups of the used nuclear fuel will increase. Thus, it is necessary to obtain data on high burnup fuel to demonstrate that the used nuclear fuel remains intact after extended storage. An evaluation was performed to determine the conditions that may lead to failure of dry storage SSC's. This paper documents the initial technical gap analysis performed to identify data and modeling needs to develop the desired technical bases to ensure the safety functions of dry stored fuel.

  16. Summary Report for Capsule Dry Storage Project

    SciTech Connect (OSTI)

    JOSEPHSON, W S

    2003-09-04T23:59:59.000Z

    There are 1.936 cesium (Cs) and strontium (Sr) capsules stored in pools at the Waste Encapsulation and Storage Facility (WESF). These capsules will be moved to dry storage on the Hanford Site as an interim measure to reduce risk. The Cs/Sr Capsule Dry Storage Project (CDSP) is conducted under the assumption the capsules will eventually be moved to the repository at Yucca Mountain, and the design criteria include requirements that will facilitate acceptance at the repository. The storage system must also permit retrieval of capsules in the event vitrification of the capsule contents is pursued. A cut away drawing of a typical cesium chloride (CsCI) capsule and the capsule property and geometry information are provided in Figure 1.1. Strontium fluoride (SrF{sub 2}) capsules are similar in design to CsCl capsules. Further details of capsule design, current state, and reference information are given later in this report and its references. Capsule production and life history is covered in WMP-16938, Capsule Characterization Report for Capsule Dry Storage Project, and is briefly summarized in Section 5.2 of this report.

  17. Shielding analysis of the NAC-MPC storage system

    SciTech Connect (OSTI)

    Napolitano, D.G.; Romano, N.J. [NAC International, Norcross, GA (United States); Hertel, N.E. [Georgia Institute of Technology, Atlanta, GA (United States)] [and others

    1997-12-01T23:59:59.000Z

    This paper presents the shielding analyses of the NAC-MPC dry cask storage system. The NAC-MPC dry cask storage system consists of a transportable storage canister, a transfer cask, and a vertical concrete storage cask. The NAC-MPC is designed to accommodate 36 {open_quotes}Yankee Class{close_quotes} fuel assemblies with a maximum burnup of 36,000 MWd/tonne U burnup and 8 yr cooling time. The shielding analysis is performed with the SCALE 4.3 code package which includes SAS2H for source term generation and SAS4A, a modification of SAS4, for shielding evaluations. SAS4 utilizes a one-dimensional XSDRNPM adjoint calculation of the cask to generate biasing parameters for a three-dimensional MORSE-SGC Monte Carlo model of the cask geometry.

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

    SciTech Connect (OSTI)

    Nichol, M. [Nuclear Energy Institute, Washington DC (United States)

    2012-07-01T23:59:59.000Z

    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 these efforts to redefine retrievability to apply to the dual purpose cask, and not to apply to individual assemblies. (authors)

  19. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    SciTech Connect (OSTI)

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20T23:59:59.000Z

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  20. Heat transfer modeling of dry spent nuclear fuel storage facilities

    SciTech Connect (OSTI)

    Lee, S.Y.

    1999-07-01T23:59:59.000Z

    The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geologic codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geologic repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.

  1. Heat Transfer Modeling of Dry Spent Nuclear Fuel Storage Facilities

    SciTech Connect (OSTI)

    Lee, S.Y.

    1999-01-13T23:59:59.000Z

    The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geological codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geological repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.

  2. Development of a conditioning system for the dual-purpose transport and storage cask for spent nuclear fuel from decommissioned Russian submarines

    SciTech Connect (OSTI)

    Dyer, R.S.; Barnes, E. [U.S. Environmental Protection Agency, Washington D.C. (United States); Snipes, R.L. [Oak Ridge National Laboratory, TN (United States); Guskov, V.; Makarchuk, T. [Special Mechanical Engineering Design Bureau (KBSM), St. Petersburg (Russian Federation)

    2007-07-01T23:59:59.000Z

    Russia, stores large quantities of spent nuclear fuel (SNF) from submarine and ice-breaker nuclear powered naval vessels. This high-level radioactive material presents a significant threat to the Arctic and marine environments. Much of the SNF from decommissioned Russian nuclear submarines is stored either onboard the submarines or in floating storage vessels in Northwest and Far East Russia. Some of the SNF is damaged, stored in an unstable condition, or of a type that cannot currently be reprocessed. In many cases, the existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing all of this fuel from remote locations. Additional transport and storage options are required. Some of the existing storage facilities being used in Russia do not meet health and safety and physical security requirements. The U.S. has assisted Russia in the development of a new dual-purpose metal-concrete transport and storage cask (TUK-108/1) for their military SNF and assisted them in building several new facilities for off-loading submarine SNF and storing these TUK-108/1 casks. These efforts have reduced the technical, ecological, and security challenges for removal, handling, interim storage, and shipment of this submarine fuel. Currently, Russian licensing limits the storage period of the TUK-108/1 casks to no more than two years before the fuel must be shipped for reprocessing. In order to extend this licensed storage period, a system is required to condition the casks by removing residual water and creating an inert storage environment by backfilling the internal canisters with a noble gas such as argon. The U.S. has assisted Russia in the development of a mobile cask conditioning system for the TUK-108/1 cask. This new conditioning system allows the TUK 108/1 casks to be stored for up to five years after which the license may be considered for renewal for an additional five years or the fuel will be shipped to 'Mayak' for reprocessing. The U.S. Environmental Protection Agency (EPA), in cooperation with the U.S. DOD Office of Cooperative Threat Reduction (CTR), and the DOE's ORNL, along with the Norwegian Defense Research Establishment, worked closely with the Ministry of Defense and the Ministry of Atomic Energy of the Russian Federation (RF) to develop an improved integrated management system for interim storage of military SNF in Russia. The initial Project activities included: (1) development of a prototype dual-purpose, metal-concrete 40-ton cask for both the transport and interim storage of RF SNF, and (2) development of the first transshipment/interim storage facility for these casks in Murmansk. The U.S. has continued support to the project by assisting the RF with the development of the first mobile system that provides internal conditioning for the TUK-108/1 casks to allow them to be stored for longer than the current licensing period of two years. Development of the prototype TUK-108/1 cask was completed in December 2000 under the Arctic Military Environmental Cooperation (AMEC) Program. This was the first metal-concrete cask developed, licensed, and produced in the RF for both the transportation and storage of SNF from decommissioned submarines. These casks are currently being serially produced in NW Russia and 108 casks have been produced to date. Russia is using these casks for the transport and interim storage of military SNF from decommissioned nuclear submarines at naval installations in the Arctic and Far East in conformance with the Strategic Arms Reduction Treaty (START II). The design, construction, and commissioning of the first transshipment/interim storage facility in the RF was completed and ready for full operation in September 2003. Because of the RF government reorganization and changing regulations for spent fuel storage facilities, the storage facility at Murmansk was not fully licensed for operation until December 2005. The RF has reported that the facility is now fully operational. The TUK-108/1 SNF transport and storage casks were designed

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

    SciTech Connect (OSTI)

    BAZINET, G.D.

    2000-04-13T23:59:59.000Z

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

  4. Hydrogen storage materials and method of making by dry homogenation

    DOE Patents [OSTI]

    Jensen, Craig M. (Kailua, HI); Zidan, Ragaiy A. (Honolulu, HI)

    2002-01-01T23:59:59.000Z

    Dry homogenized metal hydrides, in particular aluminum hydride compounds, as a material for reversible hydrogen storage is provided. The reversible hydrogen storage material comprises a dry homogenized material having transition metal catalytic sites on a metal aluminum hydride compound, or mixtures of metal aluminum hydride compounds. A method of making such reversible hydrogen storage materials by dry doping is also provided and comprises the steps of dry homogenizing metal hydrides by mechanical mixing, such as be crushing or ball milling a powder, of a metal aluminum hydride with a transition metal catalyst. In another aspect of the invention, a method of powering a vehicle apparatus with the reversible hydrogen storage material is provided.

  5. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    SciTech Connect (OSTI)

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01T23:59:59.000Z

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions.

  6. Thermal Analysis of a Dry Storage Concept for Capsule Dry Storage Project

    SciTech Connect (OSTI)

    JOSEPHSON, W S

    2003-09-04T23:59:59.000Z

    There are 1,936 cesium (Cs) and strontium (Sr) capsules stored in pools at the Waste Encapsulation and Storage Facility (WESF). These capsules will be moved to dry storage on the Hanford Site as an interim measure to reduce risk. The Cs/Sr Capsule Dry Storage Project is conducted under the assumption that the capsules will eventually be moved to the repository at Yucca Mountain, and the design criteria include requirements that will facilitate acceptance at the repository. The storage system must also permit retrieval of capsules in the event that vitrification of the capsule contents is pursued. The Capsule Advisory Panel (CAP) was created by the Project Manager for the Hanford Site Capsule Dry Storage Project (CDSP). The purpose of the CAP is to provide specific technical input to the CDSP; to identify design requirements; to ensure design requirements for the project are conservative and defensible; to identify and resolve emerging, critical technical issues, as requested; and to support technical reviews performed by regulatory organizations, as requested. The CAP will develop supporting and summary documents that can be used as part of the technical and safety bases for the CDSP. The purpose of capsule dry storage thermal analysis is to: (1) Summarize the pertinent thermal design requirements sent to vendors, (2) Summarize and address the assumptions that underlie those design requirements, (3) Demonstrate that an acceptable design exists that satisfies the requirements, (4) Identify key design features and phenomena that promote or impede design success, (5) Support other CAP analyses such as corrosion and integrity evaluations, and (6) Support the assessment of proposed designs. It is not the purpose of this report to optimize or fully analyze variations of postulated acceptable designs. The present evaluation will indicate the impact of various possible design features, but not systematically pursue design improvements obtainable through analysis refinements and/or relaxation of conservatisms. However, possible design improvements will be summarized for future application. All assumptions and related design features, while appropriate for conceptual designs, must be technically justified for the final design. The pertinent thermal design requirements and underlying assumptions are summarized in Section 1.3. The majority of the thermal analyses, as described in Sections 4.2 and 4.3, focus on an acceptable conceptual design arrived at by refinement of a preliminary but unacceptable design. The results of the subject thermal analyses, as presented in Section 4.0, satisfy items 3 and 4 above.

  7. Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

  8. Recommended temperature limits for dry storage of spent light water reactor Zircaloy-clad fuel rods in inert gas

    SciTech Connect (OSTI)

    Levy, I.S.; Chin, B.A.; Simonen, E.P.; Beyer, C.E.; Gilbert, E.R.; Johnson, A.B. Jr.

    1987-05-01T23:59:59.000Z

    It is concluded that the recommendation of a single-valued temperature limit of 380/sup 0/C should be replaced by multiple limits to account for variations in fuel design, burnup level, spent fuel age, and storage cask design. A single-valued limit to account for these factors would, in some situations, impose unnecessary conservatisms and, potentially, economic penalties for utilities and storage cask vendors. The technical validity and conservatism of the CSFM model should assure acceptance by the NRC for utility and cask vendor use.

  9. The Impacts of Dry-Storage Canister Designs on Spent Nuclear...

    Office of Environmental Management (EM)

    The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. The Impacts of Dry-Storage Canister Designs on Spent...

  10. Saving for dry days: Aquifer storage and recovery may help 

    E-Print Network [OSTI]

    Wythe, Kathy

    2008-01-01T23:59:59.000Z

    tx H2O | pg. 2 Saving for dry days Story by Kathy Wythe tx H2O | pg. 3 Aquifer storage and recovery may help With reoccurring droughts and growing population, Texas will always be looking for better ways to save or use water. Some water... suppliers in Texas are turning to aquifer storage and recovery. During the dry summer of 2008, the San Antonio Water System (SAWS) had enough assets in its ?bank? (of water) to make with- drawals to meet the needs of its customers. The water bank...

  11. Modification and benchmarking of SKYSHINE-III for use with ISFSI cask arrays

    SciTech Connect (OSTI)

    Hertel, N.E. [Georgia Institute of Technology, Atlanta, GA (United States); Napolitano, D.G. [NAC International, Norcross, GA (United States)

    1997-12-01T23:59:59.000Z

    Dry cask storage arrays are becoming more and more common at nuclear power plants in the United States. Title 10 of the Code of Federal Regulations, Part 72, limits doses at the controlled area boundary of these independent spent-fuel storage installations (ISFSI) to 0.25 mSv (25 mrem)/yr. The minimum controlled area boundaries of such a facility are determined by cask array dose calculations, which include direct radiation and radiation scattered by the atmosphere, also known as skyshine. NAC International (NAC) uses SKYSHINE-III to calculate the gamma-ray and neutron dose rates as a function of distance from ISFSI arrays. In this paper, we present modifications to the SKYSHINE-III that more explicitly model cask arrays. In addition, we have benchmarked the radiation transport methods used in SKYSHINE-III against {sup 60}Co gamma-ray experiments and MCNP neutron calculations.

  12. Storage capacity in hot dry rock reservoirs

    DOE Patents [OSTI]

    Brown, Donald W. (Los Alamos, NM)

    1997-01-01T23:59:59.000Z

    A method of extracting thermal energy, in a cyclic manner, from geologic strata which may be termed hot dry rock. A reservoir comprised of hot fractured rock is established and water or other liquid is passed through the reservoir. The water is heated by the hot rock, recovered from the reservoir, cooled by extraction of heat by means of heat exchange apparatus on the surface, and then re-injected into the reservoir to be heated again. Water is added to the reservoir by means of an injection well and recovered from the reservoir by means of a production well. Water is continuously provided to the reservoir and continuously withdrawn from the reservoir at two different flow rates, a base rate and a peak rate. Increasing water flow from the base rate to the peak rate is accomplished by rapidly decreasing backpressure at the outlet of the production well in order to meet periodic needs for amounts of thermal energy greater than a baseload amount, such as to generate additional electric power to meet peak demands. The rate of flow of water provided to the hot dry rock reservoir is maintained at a value effective to prevent depletion of the liquid

  13. Storage capacity in hot dry rock reservoirs

    DOE Patents [OSTI]

    Brown, D.W.

    1997-11-11T23:59:59.000Z

    A method is described for extracting thermal energy, in a cyclic manner, from geologic strata which may be termed hot dry rock. A reservoir comprised of hot fractured rock is established and water or other liquid is passed through the reservoir. The water is heated by the hot rock, recovered from the reservoir, cooled by extraction of heat by means of heat exchange apparatus on the surface, and then re-injected into the reservoir to be heated again. Water is added to the reservoir by means of an injection well and recovered from the reservoir by means of a production well. Water is continuously provided to the reservoir and continuously withdrawn from the reservoir at two different flow rates, a base rate and a peak rate. Increasing water flow from the base rate to the peak rate is accomplished by rapidly decreasing backpressure at the outlet of the production well in order to meet periodic needs for amounts of thermal energy greater than a baseload amount, such as to generate additional electric power to meet peak demands. The rate of flow of water provided to the hot dry rock reservoir is maintained at a value effective to prevent depletion of the liquid inventory of the reservoir. 4 figs.

  14. Saving for dry days: Aquifer storage and recovery may help

    E-Print Network [OSTI]

    Wythe, Kathy

    2008-01-01T23:59:59.000Z

    underground storage (MUS) of recoverable water. The Committee on Sustainable Underground Storage of Recoverable Water uses MUS ?to denote purposeful recharge of water into an aquifer system for intended recovery and use as an element of long-term water...tx H2O | pg. 2 Saving for dry days Story by Kathy Wythe tx H2O | pg. 3 Aquifer storage and recovery may help With reoccurring droughts and growing population, Texas will always be looking for better ways to save or use water. Some water...

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

    SciTech Connect (OSTI)

    Edwards, W.S.

    1997-07-14T23:59:59.000Z

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  16. Evaluation of the Technical Basis for Extended Dry Storage and

    E-Print Network [OSTI]

    -- Executive Summary U.S. Nuclear Waste Technical Review Board December 2010 #12;U.S.U.S. Nuclear Waste Technical Review Board Authors This report was prepared for the U.S. Nuclear Waste Technical Review Board.NWTRB.GOV ii #12;Extended Dry Storage and Transportation of Used Nuclear Fuel U.S. Nuclear Waste Technical

  17. Viability of Existing INL Facilities for Dry Storage Cask Handling R1 |

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group current C3EDepartmentDepartment(GATE) |DepartmentandJanuary 22,

  18. High Burnup Dry Storage Cask Research and Development Project: Final Test

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy ChinaofSchaefer To:Department of Energy CompletingPresented By: WALTER|name_firstPlan

  19. Life cycle cost report of VHLW cask

    SciTech Connect (OSTI)

    NONE

    1995-06-01T23:59:59.000Z

    This document, the Life Cycle Cost Report (LCCR) for the VHLW Cask, presents the life cycle costs for acquiring, using, and disposing of the VHLW casks. The VHLW cask consists of a ductile iron cask body, called the shielding insert, which is used for storage and transportation, and ultimately for disposal of Defense High Level Waste which has been vitrified and placed into VHLW canisters. Each ductile iron VHLW shielding insert holds one VHLW canister. For transportation, the shielding insert is placed into a containment overpack. The VHLW cask as configured for transportation is a legal weight truck cask which will be licensed by NRC. The purpose of this LCCR is to present the development of the life cycle costs for using the VHLW cask to transport VHLW canisters from the generating sites to a disposal site. Life cycle costs include the cost of acquiring, operating, maintaining, and ultimately dispositioning the VHLW cask and its associated hardware. This report summarizes costs associated with transportation of the VHLW casks. Costs are developed on the basis of expected usage, anticipated source and destination locations, and expected quantities of VHLW which must be transported. DOE overhead costs, such as the costs associated with source and destination facility handling of the VHLW, are not included. Also not included are costs exclusive to storage or disposal of the VHLW waste.

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

    SciTech Connect (OSTI)

    CARRELL, R D

    2002-07-16T23:59:59.000Z

    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.

  1. Dry Transfer Systems for Used Nuclear Fuel

    SciTech Connect (OSTI)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01T23:59:59.000Z

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  2. MODELING HEAT TRANSFER IN SPENT FUEL TRANSFER CASK NEUTRON SHIELDS – A CHALLENGING PROBLEM IN NATURAL CONVECTION

    SciTech Connect (OSTI)

    Fort, James A.; Cuta, Judith M.; Bajwa, C.; Baglietto, E.

    2010-07-18T23:59:59.000Z

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10-15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.

  3. A GAMMA RAY SCANNING APPROACH TO QUANTIFY SPENT FUEL CASK RADIONUCLIDE CONTENTS

    SciTech Connect (OSTI)

    Branney, S.

    2011-07-01T23:59:59.000Z

    The International Atomic Energy Agency (IAEA) has outlined a need to develop methods of allowing re-verification of LWR spent fuel stored in dry storage casks without the need of a reference baseline measurement. Some scanning methods have been developed, but improvements can be made to readily provide required data for spent fuel cask verification. The scanning process should be conditioned to both confirm the contents and detect any changes due to container/contents degradation or unauthorized removal or tampering. Savannah River National Laboratory and The University of Tennessee are exploring a new method of engineering a high efficiency, cost effective detection system, capable of meeting the above defined requirements in a variety of environmental situations. An array of NaI(Tl) detectors, arranged to form a 'line scan' along with a matching array of 'honeycomb' collimators provide a precisely defined field of view with minimal degradation of intrinsic detection efficiency and with significant scatter rejection. Scanning methods are adapted to net optimum detection efficiency of the combined system. In this work, and with differing detectors, a series of experimental demonstrations are performed that map system spatial performance and counting capability before actual spent fuel cask scans are performed. The data are evaluated to demonstrate the prompt ability to identify missing fuel rods or other content abnormalities. To also record and assess cask tampering, the cask is externally examined utilizing FTIR hyper spectral and other imaging/sensing approaches. This provides dated records and indications of external abnormalities (surface deposits, smears, contaminants, corrosion) attributable to normal degradation or to tampering. This paper will describe the actual gathering of data in both an experimental climate and from an actual spent fuel dry storage cask, and how an evaluation may be performed by an IAEA facility inspector attempting to draw an independent safeguards conclusion concerning the status of the special nuclear material.

  4. Effect of residual stress on the life prediction of dry storage canisters for used nuclear fuel

    E-Print Network [OSTI]

    Black, Bradley P. (Bradley Patrick)

    2013-01-01T23:59:59.000Z

    Used nuclear fuel dry storage canisters will likely be tasked with holding used nuclear fuel for a period longer than originally intended. Originally designed for 20 years, the storage time will likely approach 100 years. ...

  5. Instrumented, Shielded Test Canister System for Evaluation of Spent Nuclear Fuel in Dry Storage

    SciTech Connect (OSTI)

    Sindelar, R.L.

    1999-10-21T23:59:59.000Z

    This document describes the development of an instrumented, shielded test canister system to store and monitor aluminum-based spent nuclear duel under dry storage conditions.

  6. Studies and research concerning BNFP: cask handling equipment standardization

    SciTech Connect (OSTI)

    McCreery, Paul N.

    1980-10-01T23:59:59.000Z

    This report covers the activities of one of the sub-tasks within the Spent LWR Fuel Transportation Receiving, Handling, and Storage program. The sub-task is identified as Cask Handling Equipment Standardization. The objective of the sub-task specifies: investigate and identify opportunities for standardization of cask interface equipment. This study will examine the potential benefits of standardized yokes, decontamination barriers and special tools, and, to the extent feasible, standardized methods and software for handling the variety of casks presently available in the US fleet. The result of the investigations is a compilation of reports that are related by their common goal of reducing cask turnaround time.

  7. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R. [and others

    1996-07-01T23:59:59.000Z

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  8. A 2-D Test Problem for CFD Modeling Heat Transfer in Spent Fuel Transfer Cask Neutron Shields

    SciTech Connect (OSTI)

    Zigh, Ghani; Solis, Jorge; Fort, James A.

    2011-01-14T23:59:59.000Z

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 5-10 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper presents results for a simple 2-D problem that is an effective numerical analog for the neutron shield application. Because it is 2-D, solutions can be obtained relatively quickly allowing a comparison and assessment of sensitivity to model parameter changes. Turbulence models are considered as well as the tradeoff between steady state and transient solutions. Solutions are compared for two commercial CFD codes, FLUENT and STAR-CCM+. The results can be used to provide input to the CFD Best Practices for this application. Following study results for the 2-D test problem, a comparison of simulation results is provided for a high Rayleigh number experiment with large annular gap. Because the geometry of this validation is significantly different from the neutron shield, and due to the critical nature of this application, the argument is made for new experiments at representative scales

  9. A cask maintenance facility feasibility study

    SciTech Connect (OSTI)

    Rennich, M.J.; Medley, L.G.; Attaway, C.R.

    1989-01-01T23:59:59.000Z

    The Oak Ridge National Laboratory (ORNL) is developing a transportation system for spent nuclear fuel (SNF) and defense high level waste (HLW) as a part of the Federal Waste Management System (FWMS). In early 1988, a feasibility study was undertaken to design a stand-alone, ''green field'' facility for maintaining the FWMS casks. The feasibility study provided an initial layout facility design, an estimate of the construction cost, and an acquisition schedule for a Cask Maintenance Facility (CMF). The study also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrievable Storage (MRS) facility. The data, design, and estimated costs resulting from the study have been organized for use in the total transportation system decision-making process. Most importantly, the feasibility study also provides a foundation for continuing design and planning efforts. Fleet servicing facility studies, operational studies from current cask system operators, a definition of the CMF system requirements, and the experience of others in the radioactive waste transportation field were used as a basis for the feasibility study. In addition, several cask handling facilities were visited to observe and discuss cask operations to establish the functions and methods of cask maintenance expected to be used in the facility. Finally, a peer review meeting was held at Oak Ridge, Tennessee in August, 1988, in which the assumptions, design, layout, and functions of the CMF were significantly refined. Attendees included representatives from industry, the repository and transportation operations.

  10. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect (OSTI)

    Samuel Bays; Ayodeji Alajo

    2010-05-01T23:59:59.000Z

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  11. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01T23:59:59.000Z

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  12. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01T23:59:59.000Z

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  13. Thermodynamic Investigations of Aqueous Ternary Complexes for Am/Cm Separation

    E-Print Network [OSTI]

    Leggett, Christina Joy

    2012-01-01T23:59:59.000Z

    Commission: Fact Sheet: Dry Cask Storage of Spent Nucleardoc-collections/fact-sheets/dry-cask-storage.pdf (accessed

  14. A cask fleet operations study

    SciTech Connect (OSTI)

    Not Available

    1988-03-01T23:59:59.000Z

    This document describes the cask fleet currently available to transport spent nuclear fuels. The report describes the proposed operational procedures for these casks and the vehicles intended to transport them. Included are techniques for loading the cask, lifting it onto the transport vehicle, preparing the invoices, and unloading the cask at the destination. The document concludes with a discussion on the maintenance and repair of the casks. (tem) 29 figs.

  15. FACSIM/MRS-1: Cask receiving and consolidation performance assessment

    SciTech Connect (OSTI)

    Lotz, T.L.; Shay, M.R.

    1987-06-01T23:59:59.000Z

    A simulation analysis was completed to assess the performance of the shipping cask receiving and spent-fuel handling, consolidation and canistering operations of the Monitored Retrievable Storage (MRS) facility. One purpose of this evaluation was to estimate the limits of MRS operational capabilities and factors leading to those limitations. The model used to obtain the performance assessment, FACSIM/MRS-1, is one of two components of the FACSIM model developed by PNL's simulation effort for the nuclear waste-handling facility. FACSIM/MRS-1 provides the user with information about lag-storage requirements, machine use, cask queues, welder queues, and cask process and cask turnaround times. The model can help determine the effect that the following activities have on operating efficiency: (1) receiving multiple cask shipments, when rail-cask or truck-cask shipments arrive at the facility in groups of two or more, and (2) operating the facility five days per week, three shifts per day or seven days per week, three shifts per day for any conditions. In addition, sensitivity to equipment failure frequency and the time needed for equipment repair can be studied. Information on the above operating characteristics may be obtained for any spent-fuel rate, any split of shipments between truck and rail transport, or any split of boiling water reactor/pressurized water reactor fuel.

  16. Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

    SciTech Connect (OSTI)

    McConnell, J.W., Jr; Ayers, A.L. Jr; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1996-02-01T23:59:59.000Z

    This report provides a graded approach for classification of components used in transportation packaging and dry spent fuel storage systems. This approach provides a method for identifying, the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements The general types of transportation packagings and dry storage systems were identified. Discussions were held with suppliers and fabricators of packagings and storage systems to determine current practices. The methodology used in this report is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This report also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems. The safety importance of each component is discussed, and a classification category is assigned.

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

    SciTech Connect (OSTI)

    Unterzuber, R.; Wright, J.B.

    1980-09-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Oberson, Greg; Dunn, Darrell [U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington DC, 20555 (United States)] [U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington DC, 20555 (United States); Mintz, Todd; He, Xihua; Pabalan, Roberto; Miller, Larry [Center for Nuclear Waste Regulatory Analyses, 6220 Culebra Rd, San Antonio TX, 78238 (United States)] [Center for Nuclear Waste Regulatory Analyses, 6220 Culebra Rd, San Antonio TX, 78238 (United States)

    2013-07-01T23:59:59.000Z

    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 to date indicate that the deliquescence RH for sea salt is close to that of MgCl{sub 2} pure salt. SCC is observed between 35 and 80 deg. C when the ambient (RH) is close to or higher than this level, even for a low surface salt concentration. (authors)

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

    SciTech Connect (OSTI)

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

    1982-09-01T23:59:59.000Z

    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).

  20. Modular vault dry storage at Paks NPP technology and experience

    SciTech Connect (OSTI)

    Bower, C.C.F. [Gec Alsthom Engineering Systems, Leicester (United Kingdom); Szabo, B. [Paks NPP (Hungary)

    1995-12-31T23:59:59.000Z

    Paks NPP in Hungary, with its four VVER440 reactors, generates 50% of Hungary`s electricity. In 1990, it was faced with an uncertain future due to the changing political situation in Eastern Europe. The fuel storage ponds were rapidly filling up, with no secure route for disposal. The paper outlines the Paks approach to resolving the problem and the background to its chosen solution, concluding with a review of the experience of other applications of the system.

  1. MCO loading and cask loadout technical manual

    SciTech Connect (OSTI)

    PRAGA, A.N.

    1998-10-01T23:59:59.000Z

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

  2. Improvement of operational safety of dual-purpose transport packaging set for naval SNF in storage

    SciTech Connect (OSTI)

    Guskov, Vladimir; Korotkov, Gennady [JSC 'KBSM' (Russian Federation); Barnes, Ella [US Environmental Protection Agency - EPA (United States); Snipes, Randy [Oak Ridge National Laboratory - ORNL, 1 Bethel Valley Rd, Oak Ridge, TN 37830 (United States)

    2007-07-01T23:59:59.000Z

    Available in abstract form only. Full text of publication follows: In recent ten years a new technology of management of irradiated nuclear fuel (SNF) at the final stage of fuel cycle has been intensely developing on a basis of a new type of casks used for interim storage of SNF and subsequent transportation therein to the place of processing, further storage or final disposal. This technology stems from the concept of a protective cask which provides preservation of its content (SNF) and fulfillment of all other safety requirements for storage and transportation of SNF. Radiation protection against emissions and non-distribution of activity outside the cask is ensured by physical barriers, i.e. all-metal or composite body, shells, inner cavities for irradiated fuel assemblies (SFA), lids with sealing systems. Residual heat release of SFA is discharged to the environment by natural way: through emission and convection of surrounding air. By now more than 100 dual purpose packaging sets TUK-108/1 are in operation in the mode of interim storage and transportation of SNF from decommissioned nuclear powered submarines (NPS). In accordance with certificate, spent fuel is stored in TUK-108/1 on the premises of plants involved in NPS dismantlement for 2 years, whereupon it is transported for processing to PO Mayak. At one Far Eastern plant Zvezda involved in NPS dismantlement there arose a complicated situation due to necessity to extend period of storage of SNF in TUK- 108/1. To ensure safety over a longer period of storage of SNF in TUK-108/1 it is essential to modify conditions of storage by removing of residual water and filling the inner cavity of the cask with an inert gas. Within implementation of the international 1.1- 2 project Development of drying technology for the cask TUK-108/1 intended for naval SNF under the Program, there has been developed the technology of preparation of the cask for long-term storage of SNF in TUK-108/1, the design of a mobile TUK-108/1 drying facility; a pilot facility has been manufactured. This report describes key issues of cask drying technology, justification of terms of dry storage of naval SNF in no-108/1, design features of the mobile drying facility, results of tests of the pilot facility at the Far Eastern plant Zvezda. (authors)

  3. Assessment of the Fingerprinting Method for Spent Fuel Verification in MACSTOR KN-400 CANDU Dry Storage 

    E-Print Network [OSTI]

    Gowthahalli Chandregowda, Nandan

    2012-10-19T23:59:59.000Z

    is necessary in order for the International Atomic Energy Agency (IAEA) to meet with safeguards regulations. The IAEA is interested in having a new effective method of re-verification of the nuclear material in the MACSTOR KN-400 dry storage facility...

  4. Beneficial Uses Shipping System cask

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    The Beneficial Uses Shipping System (BUSS) cask is a Type B packaging currently under development at Sandia National Laboratories for the US Department of Energy (DOE). The cask will transport radioactive source capsules (CsCl and SrF2) to facilities such as sewage, food, and medical products irradiators. This paper describes the cask system and the analyses performed to predict the response of the cask in impact, puncture, and fire accident conditions as specified in the regulations. The cask prototype is being fabricated, and activities to obtain a certificate of compliance have been initiated.

  5. Opportunities to increase the productivity of spent fuel shipping casks in the United States

    SciTech Connect (OSTI)

    Winsor, G.H.; Faletti, D.W.; DeSteese, J.G.

    1980-03-01T23:59:59.000Z

    Trends indicate that future transportation requirements for spent fuel will be different from those anticipated when the current generation of casks and vehicles was designed. Increased storage capacity at most reactors will increase the average post irradiation age of the spent fuel to be transported. A scenario is presented which shows the 18 casks currently available should be sufficient until approximately 1983. Beyond this time, it appears that an adequate transportation system can be maintained by acquiring, as needed, casks of current designs and new casks currently under development. Spent fuel transportation requirements in the post-1990 period can be met by a new generation of casks specifically designed to transport long-cooled fuel. In terms of the number of casks needed, productivity may be increased by 19% if rail cask turnaround time is reduced to 4 days from the current range of 6.5 to 8.5 days. Productivity defined as payloads per cask year could be increased 62% if the turnaround time for legal weight truck casks were reduced from 12 hours to 4 hours. On a similar basis, overweight truck casks show a 28% increase in productivity.

  6. Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

    SciTech Connect (OSTI)

    Koji Shirai

    2006-04-01T23:59:59.000Z

    The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

  7. Re-evaluation of monitored retrievable storage concepts

    SciTech Connect (OSTI)

    Fletcher, J.F.; Smith, R.I.

    1989-04-01T23:59:59.000Z

    In 1983, as a prelude to the monitored retrievable storage (MRS) facility conceptual design, the Pacific Northwest Laboratory (PNL) conducted an evaluation for the US Department of Energy (DOE) that examined alternative concepts for storing spent LWR fuel and high- level wastes from fuel reprocessing. The evaluation was made considering nine concepts for dry away-from-reactor storage. The nine concepts evaluated were: concrete storage cask, tunnel drywell, concrete cask-in-trench, open-cycle vault, metal casks (transportable and stationary), closed-cycle vault, field drywell, and tunnel-rack vault. The purpose and scope of the re-evaluation did not require a repetition of the expert-based examinations used earlier. Instead, it was based on more detailed technical review by a small group, focusing on changes that had occurred since the initial evaluation was made. Two additional storage concepts--the water pool and the horizontal modular storage vault (NUHOMS system)--were ranked along with the original nine. The original nine concepts and the added two conceptual designs were modified as appropriate for a scenario with storage capacity for 15,000 MTU of spent fuel. Costs, area requirements, and technical and historical data pertaining to MRS storage were updated for each concept.

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

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Pardini, Allan F.; Cuta, Judith M.; Adkins, Harold E.; Casella, Andrew M.; Qiao, Hong (Amy); Larche, Michael R.; Diaz, Aaron A.; Doctor, Steven R.

    2013-09-01T23:59:59.000Z

    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.

  9. Spent Nuclear Fuel Dry Transfer System Cold Demonstration Project Final Report

    SciTech Connect (OSTI)

    Christensen, Max R; McKinnon, M. A.

    1999-12-01T23:59:59.000Z

    The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, a number of DTS prototype systems were modified to apply lessons learned to date. Additional testing was performed to validate the modifications. In general, all the equipment worked exceptionally well. The demonstration also helped confirm cost estimates that had been made at several points in the development of the system.

  10. Profits and Losses from On-farm Drying and Storage of Grain Sorghum in Central Texas and the Coastal Bend.

    E-Print Network [OSTI]

    Hildreth, R. J.; Moore, C. A.

    1958-01-01T23:59:59.000Z

    DEPARTMENT OF AGRICULTURE SUMMARY The cost of owning and operating round bins and drying equipment when used at capacity for on-farm drying and storage of grain sorghuni in the Coastal Bend area was 34 cents per hundred- weight and 30 cents per... hundredweight in the Central Texas area; These costs- were based on a study of 91 units over two drying and storage seasons, 1954-55 and 1955-56. The costs with a building were slightly higher. The seasonal price spread cannot be compared directly...

  11. COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA

    SciTech Connect (OSTI)

    Dyer, R.S.; Barnes, E.; Snipes, R.L.; Hoeibraaten, S.; Gran, H.C.; Foshaug, E.; Godunov, V.

    2003-02-27T23:59:59.000Z

    Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal.

  12. Profits and Losses from On-farm Drying and Storage of Rice in Texas.

    E-Print Network [OSTI]

    Sorenson, J.W. Jr.; Hildreth, R.J.

    1957-01-01T23:59:59.000Z

    , depending on the type of unit. The total cost for the round bin with a portable auger was 66 cents per barrel; a building with an air conveyor, 90 cents per barrel; a building with an installed auger, 79 cents per barrel; and a building with a portable... auger, 74 cents. A comparison of the average benefits from drying and storage, $2.17 per barrel, with the total cost of the most common type unit, a build- ing with an installed auger of 79 cents per barrel, indicates a profit of $1.38 per barrel...

  13. CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    E.F. Loros

    2000-06-23T23:59:59.000Z

    The Carrier/Cask Handling System receives casks on railcars and legal-weight trucks (LWTs) (transporters) that transport loaded casks and empty overpacks to the Monitored Geologic Repository (MGR) from the Carrier/Cask Transport System. Casks that come to the MGR on heavy-haul trucks (HHTs) are transferred onto railcars before being brought into the Carrier/Cask Handling System. The system is the interfacing system between the railcars and LWTs and the Assembly Transfer System (ATS) and Canister Transfer System (CTS). The Carrier/Cask Handling System removes loaded casks from the cask transporters and transfers the casks to a transfer cart for either the ATS or CTS, as appropriate, based on cask contents. The Carrier/Cask Handling System receives the returned empty casks from the ATS and CTS and mounts the casks back onto the transporters for reshipment. If necessary, the Carrier/Cask Handling System can also mount loaded casks back onto the transporters and remove empty casks from the transporters. The Carrier/Cask Handling System receives overpacks from the ATS loaded with canisters that have been cut open and emptied and mounts the overpacks back onto the transporters for disposal. If necessary, the Carrier/Cask Handling System can also mount empty overpacks back onto the transporters and remove loaded overpacks from them. The Carrier/Cask Handling System is located within the Carrier Bay of the Waste Handling Building System. The system consists of cranes, hoists, manipulators, and supporting equipment. The Carrier/Cask Handling System is designed with the tooling and fixtures necessary for handling a variety of casks. The Carrier/Cask Handling System performance and reliability are sufficient to support the shipping and emplacement schedules for the MGR. The Carrier/Cask Handling System interfaces with the Carrier/Cask Transport System, ATS, and CTS as noted above. The Carrier/Cask Handling System interfaces with the Waste Handling Building System for building structures and space allocations. The Carrier/Cask Handling System interfaces with the Waste Handling Building Electrical System for electrical power.

  14. Energy and Security in Northeast Asia: Proposals for Nuclear Cooperation

    E-Print Network [OSTI]

    Kaneko, Kumao; Suzuki, Atsuyuki; Choi, Jor-Shan; Fei, Edward

    1998-01-01T23:59:59.000Z

    term solutions grow. Dry cask storage could be an interimDry storage of the spent fuel is an option, but the storage casks

  15. Best Practices for Finite Element Analysis of Spent Nuclear Fuel Transfer, Storage, and Transportation Systems

    SciTech Connect (OSTI)

    Bajwa, Christopher S.; Piotter, Jason; Cuta, Judith M.; Adkins, Harold E.; Klymyshyn, Nicholas A.; Fort, James A.; Suffield, Sarah R.

    2010-08-11T23:59:59.000Z

    Storage casks and transportation packages for spent nuclear fuel (SNF) are designed to confine SNF in sealed canisters or casks, provide structural integrity during accidents, and remove decay through a storage or transportation overpack. The transfer, storage, and transportation of SNF in dry storage casks and transport packages is regulated under 10 CFR Part 72 and 10 CFR Part 71, respectively. Finite Element Analysis (FEA) is used with increasing frequency in Safety Analysis Reports and other regulatory technical evaluations related to SNF casks and packages and their associated systems. Advances in computing power have made increasingly sophisticated FEA models more feasible, and as a result, the need for careful review of such models has also increased. This paper identifies best practice recommendations that stem from recent NRC review experience. The scope covers issues common to all commercially available FEA software, and the recommendations are applicable to any FEA software package. Three specific topics are addressed: general FEA practices, issues specific to thermal analyses, and issues specific to structural analyses. General FEA practices covers appropriate documentation of the model and results, which is important for an efficient review process. The thermal analysis best practices are related to cask analysis for steady state conditions and transient scenarios. The structural analysis best practices are related to the analysis of casks and associated payload during standard handling and drop scenarios. The best practices described in this paper are intended to identify FEA modeling issues and provide insights that can help minimize associated uncertainties and errors, in order to facilitate the NRC licensing review process.

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

    SciTech Connect (OSTI)

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

    2014-01-01T23:59:59.000Z

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

  17. The impact of new short season rice varieties on drying and storage of rough rice in Texas

    E-Print Network [OSTI]

    Bhagia, Gobind Shewakram

    1967-01-01T23:59:59.000Z

    THE IMPACT OF NEW SHORT SEASON RICE VARIETIES ON DRYING AND STORAGE OF ROUGH RICE IN TEXAS A Thesis by GOBIND SHEWAKRAM BHAGIA Submitted to the Graduate College of the Texas A&M University in partial fulfillment of the requirements... for the degree of MASTER OF SCIENCE January 1967 Ma)or Sub/ect: Agricultural Economics THE IMPACT OF NEW SHORT SEASON RICE VARIETIES ON DRYING AND STORAGE OF ROUGH RICE IN TEXAS A Thesis by GOBIND SHEWAKIUiM BHAGIA Approved as to style and oontent by...

  18. Cask fleet operations study

    SciTech Connect (OSTI)

    Not Available

    1988-01-01T23:59:59.000Z

    The Nuclear Waste Policy Act of 1982 assigned to the Department of Energy's (DOE) Office of Civilian Waste Management the responsibility for disposing of high-level waste and spent fuel. A significant part of that responsibility involves transporting nuclear waste materials within the federal waste management system; that is, from the waste generator to the repository. The lead responsibility for transportation operations has been assigned to Oak Ridge Operations, with Oak Ridge National Laboratory (ORNL) providing technical support through the Transportation Operations Support Task Group. One of the ORNL support activities involves assessing what facilities, equipment and services are required to assure that an acceptable, cost-effective and safe transportation operations system can be designed, operated and maintained. This study reviews, surveys and assesses the experience of Nuclear Assurance Corporation (NAC) in operating a fleet of spent-fuel shipping casks to aid in developing the spent-fuel transportation system.

  19. FUEL CASK IMPACT LIMITER VULNERABILITIES

    SciTech Connect (OSTI)

    Leduc, D; Jeffery England, J; Roy Rothermel, R

    2009-02-09T23:59:59.000Z

    Cylindrical fuel casks often have impact limiters surrounding just the ends of the cask shaft in a typical 'dumbbell' arrangement. The primary purpose of these impact limiters is to absorb energy to reduce loads on the cask structure during impacts associated with a severe accident. Impact limiters are also credited in many packages with protecting closure seals and maintaining lower peak temperatures during fire events. For this credit to be taken in safety analyses, the impact limiter attachment system must be shown to retain the impact limiter following Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) impacts. Large casks are often certified by analysis only because of the costs associated with testing. Therefore, some cask impact limiter attachment systems have not been tested in real impacts. A recent structural analysis of the T-3 Spent Fuel Containment Cask found problems with the design of the impact limiter attachment system. Assumptions in the original Safety Analysis for Packaging (SARP) concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.

  20. CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

    SciTech Connect (OSTI)

    Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

    2005-11-07T23:59:59.000Z

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L{sub R}, for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A{sub 2}, are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate.

  1. Documentation for initial testing and inspections of Beneficial Uses Shipping System (BUSS) Cask

    SciTech Connect (OSTI)

    Lundeen, J.E.

    1994-08-25T23:59:59.000Z

    The purpose of this report is to compile data generated during the initial tests and inspections of the Beneficial Uses Shipping System (BUSS) Cask. In addition, this report will verify that the testing criteria identified in section 8.1 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The BUSS Cask body and lid are each one-piece forgings fabricated from ASTM A473, Type 304 stainless steel. The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Chapter 8 of the BUSS Cask SARP requires several acceptance tests and inspections, each intended to evaluate the performance of different components of the BUSS Cask system, to be performed before its first use. The results of the tests and inspections required are included in this document.

  2. THERMAL MODELING ANALYSIS OF SRS 70 TON CASK

    SciTech Connect (OSTI)

    Lee, S.; Jordan, J.; Hensel, S.

    2011-03-08T23:59:59.000Z

    The primary objective of this work was to perform the thermal calculations to evaluate the Material Test Reactor (MTR) fuel assembly temperatures inside the SRS 70-Ton Cask loaded with various bundle powers. MTR fuel consists of HFBR, MURR, MIT, and NIST. The MURR fuel was used to develop a bounding case since it is the fuel with the highest heat load. The results will be provided for technical input for the SRS 70 Ton Cask Onsite Safety Assessment. The calculation results show that for the SRS 70 ton dry cask with 2750 watts total heat source with a maximum bundle heat of 670 watts and 9 bundles of MURR bounding fuel, the highest fuel assembly temperatures are below about 263 C. Maximum top surface temperature of the plastic cover is about 112 C, much lower than its melting temperature 260 C. For 12 bundles of MURR bounding fuel with 2750 watts total heat and a maximum fuel bundle of 482 watts, the highest fuel assembly temperatures are bounded by the 9 bundle case. The component temperatures of the cask were calculated by a three-dimensional computational fluid dynamics approach. The modeling calculations were performed by considering daily-averaged solar heat flux.

  3. Geological Problems in Radioactive Waste Isolation: Second Worldwide Review

    E-Print Network [OSTI]

    2010-01-01T23:59:59.000Z

    centralized facility for dry cask storage of spent fuel andbeen made. Therefore, a dry in-cask storage facility was put

  4. Cask system design guidance for robotic handling

    SciTech Connect (OSTI)

    Griesmeyer, J.M.; Drotning, W.D.; Morimoto, A.K.; Bennett, P.C.

    1990-10-01T23:59:59.000Z

    Remote automated cask handling has the potential to reduce both the occupational exposure and the time required to process a nuclear waste transport cask at a handling facility. The ongoing Advanced Handling Technologies Project (AHTP) at Sandia National Laboratories is described. AHTP was initiated to explore the use of advanced robotic systems to perform cask handling operations at handling facilities for radioactive waste, and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof-of-concept robotic systems developed in AHTP are intended to extrapolate from currently available commercial systems to the systems that will be available by the time that a repository would be open for operation. The project investigates those cask handling operations that would be performed at a nuclear waste repository facility during cask receiving and handling. The ongoing AHTP indicates that design guidance, rather than design specification, is appropriate, since the requirements for robotic handling do not place severe restrictions on cask design but rather focus on attention to detail and design for limited dexterity. The cask system design features that facilitate robotic handling operations are discussed, and results obtained from AHTP design and operation experience are summarized. The application of these design considerations is illustrated by discussion of the robot systems and their operation on cask feature mock-ups used in the AHTP project. 11 refs., 11 figs.

  5. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    SciTech Connect (OSTI)

    THIELGES, J.R.; CHASTAIN, S.A.

    2007-06-21T23:59:59.000Z

    The Disposable Solid Waste Cask (DSWC) is a shielded cask capable of transporting, storing, and disposing of six non-fuel core components or approximately 27 cubic feet of radioactive solid waste. Five existing DSWCs are candidates for use in storing and disposing of non-fuel core components and radioactive solid waste from the Interim Examination and Maintenance Cell, ultimately shipping them to the 200 West Area disposal site for burial. A series of inspections, studies, analyses, and modifications were performed to ensure that these casks can be used to safely ship solid waste. These inspections, studies, analyses, and modifications are summarized and attached in this report. Visual inspection of the casks interiors provided information with respect to condition of the casks inner liners. Because water was allowed to enter the casks for varying lengths of time, condition of the cask liner pipe to bottom plate weld was of concern. Based on the visual inspection and a corrosion study, it was concluded that four of the five casks can be used from a corrosion standpoint. Only DSWC S/N-004 would need additional inspection and analysis to determine its usefulness. The five remaining DSWCs underwent some modification to prepare them for use. The existing cask lifting inserts were found to be corroded and deemed unusable. New lifting anchor bolts were installed to replace the existing anchors. Alternate lift lugs were fabricated for use with the new lifting anchor bolts. The cask tiedown frame was modified to facilitate adjustment of the cask tiedowns. As a result of the above mentioned inspections, studies, analysis, and modifications, four of the five existing casks can be used to store and transport waste from the Interim Examination and Maintenance Cell to the disposal site for burial. The fifth cask, DSWC S/N-004, would require further inspections before it could be used.

  6. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Gilbert, E.R.

    1983-09-01T23:59:59.000Z

    This report summarizes the technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. In addition, dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor (PWR) fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved {similar_to}5,000 fuel rods, and {similar_to}600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570{sup 0}C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at {similar_to}70{sup 0}C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the United States. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380{sup 0}C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400{sup 0}C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved.

  7. Nuclear Industry Input to the Development of Concepts for the Consolidated Storage of Used Nuclear Fuel - 13411

    SciTech Connect (OSTI)

    Phillips, Chris; Thomas, Ivan; McNiven, Steven [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States)] [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States); Lanthrum, Gary [NAC International, 3930 East Jones Bridge Road, Norcross, GA, 30092 (United States)] [NAC International, 3930 East Jones Bridge Road, Norcross, GA, 30092 (United States)

    2013-07-01T23:59:59.000Z

    EnergySolutions and its team partners, NAC International, Exelon Nuclear Partners, Talisman International, TerranearPMC, Booz Allen Hamilton and Sargent and Lundy, have carried out a study to develop concepts for a Consolidated Storage Facility (CSF) for the USA's stocks of commercial Used Nuclear Fuel (UNF), and the packaging and transport provisions required to move the UNF to the CSF. The UNF is currently stored at all 65 operating nuclear reactor sites in the US, and at 10 shutdown sites. The study was funded by the US Department of Energy and followed the recommendations of the Blue Ribbon Commission on America's Nuclear Future (BRC), one of which was that the US should make prompt efforts to develop one or more consolidated storage facilities for commercial UNF. The study showed that viable schemes can be devised to move all UNF and store it at a CSF, but that a range of schemes is required to accommodate the present widely varying UNF storage arrangements. Although most UNF that is currently stored at operating reactor sites is in water-filled pools, a significant amount is now dry stored in concrete casks. At the shutdown sites, the UNF is dry stored at all but two of the ten sites. Various types of UNF dry storage configurations are used at the operating sites and shutdown sites that include vertical storage casks that are also licensed for transportation, vertical casks that are licensed for storage only, and horizontally orientated storage modules. The shutdown sites have limited to nonexistent UNF handling infrastructure and several no longer have railroad connections, complicating UNF handling and transport off the site. However four methods were identified that will satisfactorily retrieve the UNF canisters within the storage casks and transport them to the CSF. The study showed that all of the issues associated with the transportation and storage of UNF from all sites in the US can be accommodated by adopting a staged approach to the construction of the CSF. Stage 1 requires only a cask storage pad and railroad interface to be constructed, and the CSF can then receive the UNF that is in transportable storage casks. Stage 2 adds a canister handling facility, a storage cask fabrication facility and an expanded storage pad, and enables the receipt of all canistered UNF from both operating and shutdown sites. Stage 3 provides a repackaging facility with a water-filled pool that provides flexibility for a range of repackaging scenarios. This includes receiving and repackaging 'bare' UNF into suitable canisters that can be placed into interim storage at the CSF, and enables UNF that is being received, or already in storage onsite, to be repackaged into canisters that are suitable for disposal at a geologic repository. The study used the 'Total System Model' (TSM) to analyze a range of CSF capacities and operating scenarios with differing parameters covering UNF pickup orders, one or more CSF sites, CSF start dates, CSF receipt rates and geologic repository start dates. The TSM was originally developed to model movement of UNF to the Yucca Mountain repository and was modified for this study to enable the CSF to become the 'gateway' to a future geologic repository. The TSM analysis enabled costs to be estimated for each scenario and showed how these are influenced by each of the parameters. This information will provide essential underpinning for a future Conceptual Design preparation. (authors)

  8. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    SciTech Connect (OSTI)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

    2012-07-01T23:59:59.000Z

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)

  9. Evaluation of drying technologies for storage and shipment of recombinant protein drug substance

    E-Print Network [OSTI]

    Vaudant, Jérôme

    2008-01-01T23:59:59.000Z

    With growing markets and increasing pipelines, biotechnology companies face a supply chain challenge to manufacture and distribute products using economically feasible methods that protect protein integrity. Adequate storage ...

  10. Safety of interim storage solutions of used nuclear fuel during extended term

    SciTech Connect (OSTI)

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M. [AREVA, 7135 Minstrel Way, Suite 300 Columbia, MD 21045 (United States)

    2013-07-01T23:59:59.000Z

    In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

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

    SciTech Connect (OSTI)

    IRWIN, J.J.

    2000-11-18T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-04-07T23:59:59.000Z

    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.

  13. Documentation for fiscal year 1995 annual BUSS cask SARP testing and inspections

    SciTech Connect (OSTI)

    Saueressig, P.T.

    1994-11-08T23:59:59.000Z

    The purpose of this report is to compile the data generated during the Fiscal Year (FY) 1995 annual tests and inspections performed on the Beneficial Uses Shipping System (BUSS) cask. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Section 8.2 ``Maintenance and Periodic Inspection Program`` of the BUSS Cask SARP requires that the following tests and inspections be performed on an annual basis: hydrostatic pressure test; helium leak test; dye penetrant test on the trunnions and life lugs; torque test on all permanent bolts; and impact limiter inspection and weight test. In addition to compiling the generated data, this report will verify that the testing criteria identified in section 8.2 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met.

  14. Capability of environmental sampling to detect undeclared cask openings

    SciTech Connect (OSTI)

    Beckstead, L.W.; Efurd, D.W.; Hemberger, P.H.; Abhold, M.E.; Eccleston, G.W.

    1997-12-01T23:59:59.000Z

    The goal of this study is to determine the signatures that would allow monitors to detect diversion of nuclear fuel (by a diverter) from a storage area such as a geological repository. Due to the complexity of operations surrounding disposal of spent nuclear fuel in a geologic repository, there are several places that a diversion of fuel could take place. After the canister that contains the fuel rods is breached, the diverter would require a hot cell to process or repackage the fuel. A reference repository and possible diversion scenarios are discussed. When a canister is breached, or during reprocessing to extract nuclear weapons material (primarily Pu), several important isotopes or signatures including tritium, {sup 85}Kr, and {sup 129}I are released to the surrounding environment and have the potential for analysis. Estimates of release concentrations of the key signatures from the repository under a hypothetical diversion scenario are presented and discussed. Gas analysis data collected from above-ground storage casks at Idaho National Engineering and Environmental Laboratory (INEEL) Test Area North (TAN) are included and discussed in the report. In addition, LANL participated in gas sampling of one TAN cask, the Castor V/21, in July 1997. Results of xenon analysis from the cask gas sample are presented and discussed. The importance of global fallout and recent commercial reprocessing activities and their effects on repository monitoring are discussed. Monitoring and instrumental equipment for analysis of the key signature isotopes are discussed along with limits of detection. A key factor in determining if diversion activities are in progress at a repository is the timeliness of detection and analysis of the signatures. Once a clandestine operation is suspected, analytical data should be collected as quickly as possible to support any evidence of diversion.

  15. Information handbook on independent spent fuel storage installations

    SciTech Connect (OSTI)

    Raddatz, M.G.; Waters, M.D.

    1996-12-01T23:59:59.000Z

    In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

  16. Radioactive materials shipping cask anticontamination enclosure

    DOE Patents [OSTI]

    Belmonte, Mark S. (Irwin, PA); Davis, James H. (Pittsburgh, PA); Williams, David A. (Pittsburgh, PA)

    1982-01-01T23:59:59.000Z

    An anticontamination device for use in storing shipping casks for radioactive materials comprising (1) a seal plate assembly; (2) a double-layer plastic bag; and (3) a water management system or means for water management.

  17. COBRA-SFS modifications and cask model optimization

    SciTech Connect (OSTI)

    Rector, D.R.; Michener, T.E.

    1989-01-01T23:59:59.000Z

    Spent-fuel storage systems are complex systems and developing a computational model for one can be a difficult task. The COBRA-SFS computer code provides many capabilities for modeling the details of these systems, but these capabilities can also allow users to specify a more complex model than necessary. This report provides important guidance to users that dramatically reduces the size of the model while maintaining the accuracy of the calculation. A series of model optimization studies was performed, based on the TN-24P spent-fuel storage cask, to determine the optimal model geometry. Expanded modeling capabilities of the code are also described. These include adding fluid shear stress terms and a detailed plenum model. The mathematical models for each code modification are described, along with the associated verification results. 22 refs., 107 figs., 7 tabs.

  18. Cask for radioactive material and method for preventing release of neutrons from radioactive material

    SciTech Connect (OSTI)

    Gaffney, M.F.; Shaffer, P.T.

    1981-09-29T23:59:59.000Z

    A cask for radioactive material, such as nuclear reactor fuel or spent nuclear reactor fuel, includes a plurality of associated walled internal compartments for containing such radioactive material, with neutron absorbing material present to absorb neutrons emitted by the radioactive material, and a plurality of thermally conductive members, such as longitudinal copper or aluminum castings, about the compartment and in thermal contact with the compartment walls and with other such thermally conductive members and having thermal contact surfaces between such members extending, preferably radially, from the compartment walls to external surfaces of the thermally conductive members, which surfaces are preferably in the form of a cylinder. The ends of the shipping cask also preferably include a neutron absorber and a conductive metal covering to dissipate heat released by decay of the radioactive material. A preferred neutron absorber utilized is boron carbide, preferably as plasma sprayed with metal powder or as particles in a matrix of phenolic polymer, and the compartment walls are preferably of stainless steel, copper or other corrosion resistant and heat conductive metal or alloy. The invention also relates to shipping casks, storage casks and other containers for radioactive materials in which a plurality of internal compartments for such material, e.g., nuclear reactor fuel rods, are joined together, preferably in modular construction with surrounding heat conductive metal members, and the modules are joined together to form a major part of a finished shipping cask, which is preferably of cylindrical shape. Also within the invention are methods of safely storing radioactive materials which emit neutrons, while dissipating the heat thereof, and of manufacturing the present shipping casks.

  19. B cell remote-handled waste shipment cask alternatives study

    SciTech Connect (OSTI)

    RIDDELLE, J.G.

    1999-05-26T23:59:59.000Z

    The decommissioning of the 324 Facility B Cell includes the onsite transport of grouted remote-handled radioactive waste from the 324 Facility to the 200 Areas for disposal. The grouted waste has been transported in the leased ATG Nuclear Services 3-82B Radioactive Waste Shipping Cask (3-82B cask). Because the 3-82B cask is a U.S. Nuclear Regulatory Commission (NRC)-certified Type B shipping cask, the lease cost is high, and the cask operations in the onsite environment may not be optimal. An alternatives study has been performed to develop cost and schedule information on alternative waste transportation systems to assist in determining which system should be used in the future. Five alternatives were identified for evaluation. These included continued lease of the 3-82B cask, fabrication of a new 3-82B cask, development and fabrication of an onsite cask, modification of the existing U.S. Department of Energy-owned cask (OH-142), and the lease of a different commercially available cask. Each alternative was compared to acceptance criteria for use in the B Cell as an initial screening. Only continued leasing of the 3-82B cask, fabrication of a new 3-82B cask, and the development and fabrication of an onsite cask were found to meet all of the B Cell acceptance criteria.

  20. THERMAL EVALUATION OF ALTERNATE SHIPPING CASK FOR GTRI EXPERIMENTS

    SciTech Connect (OSTI)

    Donna Post Guillen

    2014-06-01T23:59:59.000Z

    The Global Threat Reduction Initiative (GTRI) has many experiments yet to be irradiated in support of the High Performance Research Reactor fuels development program. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for post irradiation examination. To date, the General Electric (GE)-2000 cask has been used to transport GTRI experiments between these facilities. However, the availability of the GE-2000 cask to support future GTRI experiments is at risk. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger GTRI experiments. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled experiments. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask.

  1. Evaluation of concepts for monitored retrievable storage of spent nuclear fuel and high-level radioactive waste

    SciTech Connect (OSTI)

    Triplett, M.B.; Smith, R.I.

    1984-04-01T23:59:59.000Z

    The primary mission selected by DOE for the monitored retrieval storage (MRS) system is to provide an alternative means of storage in the event that the repository program is delayed. The MRS concepts considered were the eight concepts included in the MRS Research and Development Report to Congress (DOE 1983). These concepts are: metal cask (stationary and transportable); concrete cask (sealed storage cask); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. Conceptual design analyses were performed for the candidate concepts using a common set of design requirements specified in consideration of the MRS mission.

  2. Shock absorbing effect of the BUSS cask cooling fins

    SciTech Connect (OSTI)

    Gwinn, K.W.

    1986-06-01T23:59:59.000Z

    The structural response of the Beneficial Uses Shipping System (BUSS) cask to the hypothetical accident puncture test was determined using large deformation finite element analyses. Three orientations were considered to ensure that the most severe orientation was analyzed. These were the end, side, and center-of-gravity over corner puncture of the cask. The side puncture event, which was initially analyzed without the circumferential cooling fins, produced the most severe decelerations in the cask. Subsequent analyses, which included the cooling fins, showed a significant reduction in acceleration and yielding in the cask body. This demonstrates the viability for using cooling fins for the puncture protection of monolithic walled casks.

  3. Safety evaluation for packaging (onsite) SERF cask

    SciTech Connect (OSTI)

    Edwards, W.S.

    1997-10-24T23:59:59.000Z

    This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  4. PRESERVATION OF H2 PRODUCTION ACTIVITY IN NANOPOROUS LATEX COATINGS OF RHODOPSEUDOMONAS PALUSTRIS CGA009 DURING DRY STORAGE AT AMBIENT TEMPERATURES

    SciTech Connect (OSTI)

    Milliken, C.; Piskorska, M.; Soule, T.; Gosse, J.; Flickinger, M.; Smith, G.; Yeager, C.

    2012-08-27T23:59:59.000Z

    To assess the applicability of latex cell coatings as an "off-the-shelf' biocatalyst, the effect of osmoprotectants, temperature, humidity and O{sub 2} on preservation of H{sub 2} production in Rhodopseudomonas palustris coatings was evaluated. Immediately following latex coating coalescence (24 h) and for up to 2 weeks of dry storage, rehydrated coatings containing different osmoprotectants displayed similar rates of H{sub 2} production. Beyond 2 weeks of storage, sorbitol- treated coatings lost all H{sub 2} production activity, whereas considerable H{sub 2} production was still detected in sucrose- and trehalose-stabilized coatings. The relative humidity level at which the coatings were stored had a significant impact on the recovery and subsequent rates of H{sub 2} production. After 4 weeks storage under air at 60% humidity, coatings produced only trace amounts of H{sub 2} (0-0.1% headspace accumulation), whereas those stored at <5% humidity retained 27-53% of their H{sub 2} production activity after 8 weeks of storage. When stored in argon at <5% humidity and room temperature, R. palustris coatings retained full H{sub 2} production activity for 3 months, implicating oxidative damage as a key factor limiting coating storage. Overall, the results demonstrate that biocatalytic latex coatings are an attractive cell immobilization platform for preservation of bioactivity in the dry state.

  5. Impact Analyses and Tests of Metal Cask Considering Aircraft Engine Crash - 12308

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

    The structural integrity of a dual purpose metal cask currently under development by the Korea Radioactive Waste Management Cooperation (KRMC) is evaluated through analyses and tests under a high-speed missile impact considering the targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from the literature. The missile impact velocity was set at 150 m/s, and two impact orientations were considered. A simplified missile simulating a commercial aircraft engine is designed from an impact load history curve provided in the literature. In the analyses, the focus is on the evaluation of the containment boundary integrity of the metal cask. The analyses results are compared with the results of tests using a 1/3 scale model. The results show very good agreements, and the procedure and methodology adopted in the structural analyses are validated. While the integrity of the cask is maintained in one evaluation where the missile impacts the top side of the free standing cask, the containment boundary is breached in another case in which the missile impacts the center of the cask lid in a perpendicular orientation. A safety assessment using a numerical simulation of an aircraft engine crash into spent nuclear fuel storage systems is performed. A commercially available explicit finite element code is utilized for the dynamic simulation, and the strain rate effect is included in the modeling of the materials used in the target system and missile. The simulation results show very good agreement with the test results. It is noted that this is the first test considering an aircraft crash in Korea. (authors)

  6. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    SciTech Connect (OSTI)

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States)] [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States); Wagner, John C. [Oak Ridge National Laboratory (United States)] [Oak Ridge National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup fuel storage and transportation. This paper discusses the staff's preliminary considerations on the safety implication of fuel reconfiguration with respect to nuclear safety (subcriticality control), radiation shielding, containment, the performance of the thermal functions of the packages, and the retrievability of the contents from regulatory perspective. (authors)

  7. Operations to be Performed in the Waste Package Dry Remediation Cell

    SciTech Connect (OSTI)

    Norman E. Cole; Randy K. Elwood

    2003-10-01T23:59:59.000Z

    Describes planned and proposed operations for remediating damaged and/or out-of-compliance waste packages, casks, DPCs, overpacks, and containers at the Yucca Mountain Dry Transfer Facility.

  8. Cooking with Dry Beans

    E-Print Network [OSTI]

    Anding, Jenna

    2008-12-09T23:59:59.000Z

    This fact sheet describes the nutritonal value and safe storage of dry beans, a commodity food. It also offers food preparation ideas....

  9. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2005-04-26T23:59:59.000Z

    The purpose of this design calculation is to demonstrate that the handling operations of transportation casks performed in the Transportation Cask Receipt and Return Facility (TCRRF) and Buffer Area meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC [Bechtel SAIC Company] 2004 [DIRS 171599], Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''Transportation Cask Receipt/Return Facility Description Document'' (BSC 2004 [DIRS 170217], Section 3.2.3). Specific scope of work contained in this activity consists of the following items: (1) Evaluate criticality effects for both dry and fully flooded conditions pertaining to TCRRF and Buffer Area operations for defense in depth. (2) Evaluate Category 1 and 2 event sequences for the TCRRF as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). This evaluation includes credible fuel reconfiguration conditions. In addition to the scope of work listed above, an evaluation was also performed of modeling assumptions for commercial spent nuclear fuel (CSNF) regarding inclusion of plenum and end regions of the active fuel. This calculation is limited to CSNF and US Department of Energy (DOE) SNF. it should be mentioned that the latter waste form is evaluated more in depth in the ''Canister Handling Facility Criticality Safety Calculations (BSC 2004 [DIRS 167614]). Further, the design and safety analyses of the naval SNF canisters are the responsibility of the US 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 TCRRF and Buffer Area and may not reflect the ongoing design evolution of the facility. However, it is anticipated that design changes to the facility layout will have little or no impact on the criticality results and/or conclusions presented in this document. This calculation is subject to the ''Quality Assurance Requirements and Description'' (DOE 2004 [DIRS 171539]) because the TCRRF is included in the Q-List (BSC 2004 [DIRS 168361], p. A-3) as an item important to safety. This calculation is prepared in accordance with AP-3.12Q, ''Design Calculations and Analyses'' [DIRS 168413].

  10. Size and transportation capabilities of the existing US cask fleet

    SciTech Connect (OSTI)

    Danese, F.L. (Science Applications International Corp., Oak Ridge, TN (USA)); Johnson, P.E.; Joy, D.S. (Oak Ridge National Lab., TN (USA))

    1990-01-01T23:59:59.000Z

    This study investigates the current spent nuclear fuel cask fleet capability in the United States. In addition, it assesses the degree to which the current fleet would be available, as a contingency, until proposed Office of Civilian Radioactive Waste Management casks become operational. A limited fleet of ten spent fuel transportation casks is found to be readily available for use in Federal waste management efforts over the next decade.

  11. Transportation cask decontamination and maintenance at the potential Yucca Mountain repository; Yucca Mountain Site characterization project

    SciTech Connect (OSTI)

    Hartman, D.J.; Miller, D.D. [Bechtel National, Inc., San Francisco, CA (United States); Hill, R.R. [Sandia National Labs., Albuquerque, NM (United States)

    1992-04-01T23:59:59.000Z

    This study investigates spent fuel cask handling experience at existing nuclear facilities to determine appropriate cask decontamination and maintenance operations at the potential Yucca Mountain repository. These operations are categorized as either routine or nonroutine. Routine cask decontamination and maintenance tasks are performed in the cask preparation area at the repository. Casks are taken offline to a separate cask maintenance area for major nonroutine tasks. The study develops conceptual designs of the cask preparation area and cask maintenance area. The functions, layouts, and major features of these areas are also described.

  12. Functions of the cask maintenance facility: A white paper

    SciTech Connect (OSTI)

    Not Available

    1987-07-21T23:59:59.000Z

    The shipping cask systems are the mobile components of the transportation system, designed to safely transport spent nuclear fuel between different facilities under both normal and accident conditions. The cask system will consist of the heavily shielded cask, the cask transport vehicle (truck trailer or railcar), and any associated ancillary equipment (covers, impact limiters, lifting devices, etc.). The cask and certain parts of the cask system must be operated within the limits imposed by a certificate of compliance (COC) granted by the Nuclear Regulatory Commission (NRC). Each cask system must transport spent fuel safely during the life of the system. To maintain the operational effectiveness and safety of the cask systems, a cask maintenance facility (CMF) will be included as an integral part of the transportation system. The planning activity of the transportation system and the design effort of the CMF require that the functions to be performed by the CMF be explicitly defined. The purpose of this paper is to (1) define the potential transportation system functions to be performed at the CMF; (2) examine the impact of this functional definition on the overall transportation system; (3) identify any unresolved issues concerning the interaction of the CMF with other elements of the transportation system; and (4) make recommendations to resolve any unresolved issues so that decisions can be made early in the transportation system planning process.

  13. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    SciTech Connect (OSTI)

    none,

    1990-02-01T23:59:59.000Z

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.

  14. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect (OSTI)

    Forsberg, Charles W.; Dole, Leslie R. [Nuclear Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN, 37831-6165 (United States)

    2007-07-01T23:59:59.000Z

    Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can out-perform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal conductivity cask. A multi-year, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports. (authors)

  15. Dose Rate Analysis Capability for Actual Spent Fuel Transportation Cask Contents

    SciTech Connect (OSTI)

    Radulescu, Georgeta [ORNL] [ORNL; Lefebvre, Robert A [ORNL] [ORNL; Peplow, Douglas E. [ORNL] [ORNL; Williams, Mark L [ORNL] [ORNL; Scaglione, John M [ORNL] [ORNL

    2014-01-01T23:59:59.000Z

    The approved contents for a U.S. Nuclear Regulatory Commission (NRC) licensed spent nuclear fuel casks are typically based on bounding used nuclear fuel (UNF) characteristics. However, the contents of the UNF canisters currently in storage at independent spent fuel storage installations are considerably heterogeneous in terms of fuel assembly burnup, initial enrichment, decay time, cladding integrity, etc. Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is an integrated data and analysis system that facilitates automated cask-specific safety analyses based on actual characteristics of the as-loaded UNF. The UNF-ST&DARDS analysis capabilities have been recently expanded to include dose rate analysis of as-loaded transportation packages. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. This paper describes the UNF-ST&DARDS dose rate analysis methodology based on actual UNF canister contents and presents sample dose rate calculation results.

  16. Use of inelastic analysis in cask design

    SciTech Connect (OSTI)

    AMMERMAN,DOUGLAS J.; BREIVIK,NICOLE L.

    2000-05-15T23:59:59.000Z

    In this paper, the advantages and disadvantages of inelastic analysis are discussed. Example calculations and designs showing the implications and significance of factors affecting inelastic analysis are given. From the results described in this paper it can be seen that inelastic analysis provides an improved method for the design of casks. It can also be seen that additional code and standards work is needed to give designers guidance in the use of inelastic analysis. Development of these codes and standards is an area where there is a definite need for additional work. The authors hope that this paper will help to define the areas where that need is most acute.

  17. Nuclear cask testing films misleading and misused

    SciTech Connect (OSTI)

    Audin, L. [Audin (Lindsay), Ossining, NY (United States)

    1991-10-01T23:59:59.000Z

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  18. Nuclear cask testing films misleading and misused

    SciTech Connect (OSTI)

    Audin, L. (Audin (Lindsay), Ossining, NY (United States))

    1991-10-01T23:59:59.000Z

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  19. Feasibility study for a transportation operations system cask maintenance facility

    SciTech Connect (OSTI)

    Rennich, M.J.; Medley, L.G.; Attaway, C.R.

    1991-01-01T23:59:59.000Z

    The US Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the development of a waste management program for the disposition of spent nuclear fuel (SNF) and high-level waste (HLW). The program will include a transportation system for moving the nuclear waste from the sources to a geologic repository for permanent disposal. Specially designed casks will be used to safely transport the waste. The cask systems must be operated within limits imposed by DOE, the Nuclear Regulatory Commission (NRC), and the Department of Transportation (DOT). A dedicated facility for inspecting, testing, and maintaining the cask systems was recommended by the General Accounting Office (in 1979) as the best means of assuring their operational effectiveness and safety, as well as regulatory compliance. In November of 1987, OCRWM requested a feasibility study be made of a Cask Maintenance Facility (CMF) that would perform the required functions. 46 refs., 16 figs., 13 tabs.

  20. Development of the Beneficial Uses Shipping System cask

    SciTech Connect (OSTI)

    Yoshimura, H.R.; Wellman, G.W.; Moya, J.L.; Gonzales, A.; Gwinn, K.W.; Eakes, R.G.; Uncapher, W.L.

    1985-01-01T23:59:59.000Z

    The Beneficial Uses Shipping System (BUSS) cask is a Type B packaging currently under development at Sandia National Laboratories for the US Department of Energy (DOE) in their Beneficial Uses of Nuclear Waste Program. The cask will transport radioactive source capsules (CsCl and SrF/sub 2/) to facilities such as sewage and food irradiators. The principal design criteria for developing the BUSS cask are specified in 10CFR71. Extensive three-dimensional modeling and numerical analyses were performed to predict the effects of the impact, puncture, and fire accident conditions specified in the regulations. The cask prototype is being fabricated, and a Certificate of Compliance is being obtained. 16 references, 7 figures, 7 tables.

  1. TRANSPORTATION CASK RECEIPT AND RETURN FACILITY WORKER DOSE ASSESSMENT

    SciTech Connect (OSTI)

    V. Arakali

    2005-02-24T23:59:59.000Z

    The purpose of this design calculation is to estimate radiation doses received by personnel working in the Transportation Cask Receipt and Return Facility (TCRRF) of the repository including the personnel at the security gate and cask staging areas. This calculation is required to support the preclosure safety analysis (PCSA) to ensure that the predicted doses are within the regulatory limits prescribed by the U.S. Nuclear Regulatory Commission (NRC). The Cask Receipt and Return Facility receives NRC licensed transportation casks loaded with spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The TCRRF operation starts with the receipt, inspection, and survey of the casks at the security gate and the staging areas, and proceeds to the process facilities. The transportation casks arrive at the site via rail cars or trucks under the guidance of the national transportation system. This calculation was developed by the Environmental and Nuclear Engineering organization and is intended solely for the use of Design and Engineering in work regarding facility design. Environmental and Nuclear Engineering personnel should be consulted before using this calculation for purposes other than those stated herein or for use by individuals other than authorized personnel in the Environmental and Nuclear Engineering organization.

  2. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01T23:59:59.000Z

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM'S program specifications. (CBS)

  3. Optimization of Trajectories for the Cask and Plug Remote Handling System in Tokamak

    E-Print Network [OSTI]

    Ribeiro,Isabel

    May 2011 Optimization of Trajectories for the Cask and Plug Remote Handling System in Tokamak ­ Trajectories of the Rescue Casks · Task 4 ­ Parking in HCB Cask trajectories in level B1 of Tokamak Building System in Tokamak Building and Hot Cell o Grant Objectives · Trajectories optimization for nominal

  4. Central Storage for Unsealed Radioactive Materials

    E-Print Network [OSTI]

    Pawlowski, Wojtek

    Central Storage for Unsealed Radioactive Materials Radiation Safety Form PERMIT HOLDER NAME:______________________________ PHONE #: ____________________________ ADDRESS/DEPT.: _______________________________ Storage Location: Refrigerator Freezer Dry Storage List each item being transferred to storage separately: EH&S LAB WIPE SURVEY

  5. Farm Grain Drying and Storage 

    E-Print Network [OSTI]

    Anonymous

    CORN growing is, and is likely to remain, one of British agriculture's major enterprises. A very high proportion of the resulting harvest is now handled by combine harvesters, which bring with them problems of grain handling, ...

  6. NRC Technical Research Program to Evaluate Extended Storage and Transportation of Spent Nuclear Fuel - 12547

    SciTech Connect (OSTI)

    Einziger, R.E.; Compton, K.; Gordon, M.; Ahn, T.; Gonzales, H. [United States Nuclear Regulatory Commission, Rockville, Maryland 20852 (United States); Pan, Y. [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX 78238 (United States)

    2012-07-01T23:59:59.000Z

    Any new direction proposed for the back-end of spent nuclear fuel (SNF) cycle will require storage of SNF beyond the current licensing periods. The Nuclear Regulatory Commission (NRC) has established a technical research program to determine if any changes in the 10 CFR part 71, and 72 requirements, and associated guidance might be necessary to regulate the safety of anticipated extended storage, and subsequent transport of SNF. This three part program of: 1) analysis of knowledge gaps in the potential degradation of materials, 2) short-term research and modeling, and 3) long-term demonstration of systems, will allow the NRC to make informed regulatory changes, and determine when and if additional monitoring and inspection of the systems is necessary. The NRC has started a research program to obtain data necessary to determine if the current regulatory guidance is sufficient if interim dry storage has to be extended beyond the currently approved licensing periods. The three-phased approach consists of: - the identification and prioritization of potential degradation of the components related to the safe operation of a dry cask storage system, - short-term research to determine if the initial analysis was correct, and - a long-term prototypic demonstration project to confirm the models and results obtained in the short-term research. The gap analysis has identified issues with the SCC of the stainless steel canisters, and SNF behavior. Issues impacting the SNF and canister internal performance such as high and low temperature distributions, and drying have also been identified. Research to evaluate these issues is underway. Evaluations have been conducted to determine the relative values that various types of long-term demonstration projects might provide. These projects or follow-on work is expected to continue over the next five years. (authors)

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

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Pardini, Allan F.; Hanson, Brady D.; Sorenson, Ken B.

    2013-01-14T23:59:59.000Z

    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.

  8. Cold vacuum drying system conceptual design report

    SciTech Connect (OSTI)

    Bradshaw, F.W.

    1996-05-01T23:59:59.000Z

    This document summarizes the activities involved in the removal of the SNF from the leaking basins and to place it in stable dry storage.

  9. Cooking with Non-fat Dry Milk

    E-Print Network [OSTI]

    Anding, Jenna

    2008-12-09T23:59:59.000Z

    This fact sheet describes the nutritional value and safe storage of non-fat dry milk, a commodity food. It also offers food preparation ideas....

  10. A status report on the development and certification of the Beneficial Uses Shipping System (BUSS) cask

    SciTech Connect (OSTI)

    Yoshimura, H.R.; Bronowski, D.R.

    1996-02-01T23:59:59.000Z

    In the early 1980s, the US Department of Energy (DOE) implemented a program to encourage beneficial uses of nuclear byproduct materials, such as cesium-137 and strontium-90, created during the production of defense materials. Potential uses of the cesium-137 ({sup 137}CS) isotope included sterilizing medical products, maintaining the quality of certain food products, and disinfecting municipal sewage sludge. Strontium-90 ({sup 90}Sr) is a good heat source and has been used in thermoelectric generators and other products that require a constant supply of heat. During that same period, a proposed facility in Albuquerque, New Mexico, was designed to use cesium-137 to sterilize sewage sludge. To support the sewage sludge treatment facility, Sandia National Laboratories was funded by the DOE to develop a Nuclear Regulatory Commission (NRC)-certified Type B shipping container to transport cesium chloride (CsCl) or strontium fluoride (SrF{sub 2}) capsules produced by the Hanford Waste Encapsulation and Storage Facility (WESF) in the State of Washington. The primary purpose of the Beneficial Uses Shipping System (BUSS) cask is to provide shielding and confinement, as well as impact, puncture, and thermal protection for certified, special form contents during transport under normal and hypothetical accident conditions. The BUSS cask was designed to meet dimensional and weight constraints of the WESF and user facilities. Attaining as-low-as-reasonably-achievable (ALARA) radiation exposures in the design and operation of the transport system was a major design goal. Another goal was to obtain regulatory approval of the design by preparing a safety analysis report for packaging (SARP) (Yoshimura et al. 1993).

  11. Comparison of structural integrity of casks for spent-fuel transportation

    SciTech Connect (OSTI)

    Eifert, E.J.; Burger, C.S.; Herger, A.S. [Univ. of New Mexico, Albuquerque, NM (United States); Ammerman, D.J.; McConnell, P. [Sandia National Laboratories, Albuquerque, NM (United States)

    1995-12-31T23:59:59.000Z

    This paper presents the results of a series of finite element analyses comparing the structural integrity of two shipping casks for transportation of high-level waste (HLW). The objective of this project is to assess the advisability of utilizing ductile iron (DI) for type-B transport cask construction by investigating its structural response under severe loading conditions. This response is compared to that of a stainless steel (SS) cask under comparable loading conditions.

  12. Design analysis report for the TN-WHC cask and transportation system

    SciTech Connect (OSTI)

    Brisbin, S.A., Fluor Daniel Hanford

    1997-02-13T23:59:59.000Z

    This document presents the evaluation of the Spent Nuclear Fuel Cask and Transportation System. The system design was developed by Transnuclear, Inc. and its team members NAC International, Nelson Manufacturing, Precision Components Corporation, and Numatec, Inc. The cask is designated the TN-WHC cask. This report describes the design features and presents preliminary analyses performed to size critical dimensions of the system while meeting the requirements of the performance specification.

  13. Evaluation of the Cask Transportation Facility Modifications (CTFM) compliance to DOE order 6430.1A

    SciTech Connect (OSTI)

    ARD, K.E.

    1999-07-14T23:59:59.000Z

    This report was prepared to evaluate the compliance of Cask Transportation Facility Modifications (CTFM) to DOE Order 6430.1A.

  14. Application of the ASME code in the design of the GA-4 and GA-9 casks

    SciTech Connect (OSTI)

    Mings, W.J. (USDOE Idaho Field Office, Idaho Falls, ID (United States)); Koploy, M.A. (General Atomics, San Diego, CA (United States))

    1992-01-01T23:59:59.000Z

    General Atomics (GA) is developing two spent fuel shipping casks for transport by legal weight truck (LWT). The casks are designed to the loading, environmental conditions and safety requirements defined in Title 10 of the Code of Federal Regulations, Part 71 (10CFR71). To ensure that all components of the cask meet the 10CFR71 rules, GA established structural design criteria for each component based on NRC Regulatory Guides and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This paper discusses the criteria used for different cask components, how they were applied and the conservatism and safety margins built into the criteria and assumption.

  15. Application of the ASME code in the design of the GA-4 and GA-9 casks

    SciTech Connect (OSTI)

    Mings, W.J. [USDOE Idaho Field Office, Idaho Falls, ID (United States); Koploy, M.A. [General Atomics, San Diego, CA (United States)

    1992-08-01T23:59:59.000Z

    General Atomics (GA) is developing two spent fuel shipping casks for transport by legal weight truck (LWT). The casks are designed to the loading, environmental conditions and safety requirements defined in Title 10 of the Code of Federal Regulations, Part 71 (10CFR71). To ensure that all components of the cask meet the 10CFR71 rules, GA established structural design criteria for each component based on NRC Regulatory Guides and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This paper discusses the criteria used for different cask components, how they were applied and the conservatism and safety margins built into the criteria and assumption.

  16. Stress analysis of closure bolts for shipping casks

    SciTech Connect (OSTI)

    Mok, G.C.; Fischer, L.E. (Lawrence Livermore National Lab., CA (United States)); Hsu, S.T. (Kaiser Engineers, Oakland, CA (United States))

    1993-01-01T23:59:59.000Z

    This report specifies the requirements and criteria for stress analysis of closure bolts for shipping casks containing nuclear spent fuels or high level radioactive materials. The specification is based on existing information conceming the structural behavior, analysis, and design of bolted joints. The approach taken was to extend the ASME Boiler and Pressure Vessel Code requirements and criteria for bolting analysis of nuclear piping and pressure vessels to include the appropriate design and load characteristics of the shipping cask. The characteristics considered are large, flat, closure lids with metal-to-metal contact within the bolted joint; significant temperature and impact loads; and possible prying and bending effects. Specific formulas and procedures developed apply to the bolt stress analysis of a circular, flat, bolted closure. The report also includes critical load cases and desirable design practices for the bolted closure, an in-depth review of the structural behavior of bolted joints, and a comprehensive bibliography of current information on bolted joints.

  17. Analysis of radiation measurement data of the BUSS cask

    SciTech Connect (OSTI)

    Liu, Y.Y. [Argonne National Lab., IL (United States); Tang, J.S. [Oak Ridge National Lab., TN (United States)

    1995-12-31T23:59:59.000Z

    The Beneficial Uses Shipping System (BUSS) is a Type-B packaging developed for shipping nonfissile, special-form radioactive materials to facilities such as sewage, food, and medical-product irradiators. The primary purpose of the BUSS cask is to provide shielding and confinement, as well as impact, puncture, and thermal protection for its certified special-form contents under both normal transport and hypothetical accident conditions. A BUSS cask that contained 16 CsCl capsules (2.723 {times} 10{sup 4} TBq total activity) was recently subjected to radiation survey measurements at a Westinghouse Hanford facility, which provided data that could be used to validate computer codes. Two shielding analysis codes, MICROSHIELD (User`s Manual 1988) and SAS4 (Tan 1993), that are used at Argonne National Laboratory to evaluate the safety of packaging of radioactive materials during transportation, have been selected for analysis of radiation data obtained from the BUSS cask. MICROSHIELD, which performs only gamma radiation shielding calculation, is based on a point-kernel model with idealized geometry, whereas SAS4 is a control module in the SCALE code system (1995) that can perform three-dimensional Monte Carlo shielding calculation for photons and neutrons, with built-in procedures for cross-section data processing and automated variance reduction. The two codes differ in how they model the details of the physics of gamma photon attenuation in materials, and this difference is reflected in the associated engineering cost of the analysis. One purpose of the analysis presented in this paper, therefore, is to examine the effects of the major modeling assumptions in the two codes on calculated dose rates, and to use the measured dose rates for comparison. The focus in this paper is on analysis of radiation dose rates measured on the general body of the cask and away from penetrations.

  18. STACE: Source Term Analyses for Containment Evaluations of transport casks

    SciTech Connect (OSTI)

    Seager, K. D.; Gianoulakis, S. E. [Sandia National Labs., Albuquerque, NM (United States); Barrett, P. R.; Rashid, Y. R. [ANATECH Research Corp., La Jolla, CA (United States); Reardon, P. C. [GRAM, Inc., Albuquerque, NM (United States)

    1992-01-01T23:59:59.000Z

    Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source term has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volatile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking (e.g., the quantity and size distribution of fuel rod breaches) in which experimental validation is planned. The CRUD spallation fraction is the major area where no quantitative data has been found; therefore, this also requires experimental validation. In the interim, STACE conservatively assumes a 100% spallation fraction for computing the releasable activity. The source term methodology also conservatively assumes that there is 1 Ci of residual contamination available for release in the transport cask. However, residual contamination is still by far the smallest contributor to the source term activity.

  19. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01T23:59:59.000Z

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM`S program specifications. (CBS)

  20. Fire Hazards Analysis for the 200 Area Interim Storage Area

    SciTech Connect (OSTI)

    JOHNSON, D.M.

    2000-01-06T23:59:59.000Z

    This documents the Fire Hazards Analysis (FHA) for the 200 Area Interim Storage Area. The Interim Storage Cask, Rad-Vault, and NAC-1 Cask are analyzed for fire hazards and the 200 Area Interim Storage Area is assessed according to HNF-PRO-350 and the objectives of DOE Order 5480 7A. This FHA addresses the potential fire hazards associated with the Interim Storage Area (ISA) facility in accordance with the requirements of DOE Order 5480 7A. It is intended to assess the risk from fire to ensure there are no undue fire hazards to site personnel and the public and to ensure property damage potential from fire is within acceptable limits. This FHA will be in the form of a graded approach commensurate with the complexity of the structure or area and the associated fire hazards.

  1. Packaging Design Criteria for the MCO Cask

    SciTech Connect (OSTI)

    FLANAGAN, B.D.

    2000-08-01T23:59:59.000Z

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

  2. Status and Path Forward for the Department of Energy Used Fuel Disposition Storage and Transportation Program - 12571

    SciTech Connect (OSTI)

    Sorenson, Ken [Sandia National Laboratories (United States); Williams, Jeffrey [U.S. Department of Energy, Office of Nuclear Energy (United States)

    2012-07-01T23:59:59.000Z

    The U.S. Department of Energy, Office of Nuclear Energy (DOE/NE) has sponsored a program since Fiscal Year (FY) 09 to develop the technical basis for extended dry storage of used fuel. This program is also working to develop the transportation technical basis for the transport of used fuel after the extended storage period. As this program has progressed, data gaps associated with dry storage systems (e.g., fuel, cask internals, canister, closure, overpack, and pad) have been identified that need to be addressed to develop the technical bases for extended storage and transportation. There has also been an initiation of experimental testing and analyses based on the identified data gaps. The technical aspects of the NE program are being conducted by a multi-lab team made up of the DOE laboratories. As part of this program, a mission objective is to also collaborate closely with industry and the international sector to ensure that all the technical issues are addressed and those programs outside the DOE program can be leveraged, where possible, to maximize the global effort in storage and transportation research. The DOE/NE program is actively pursuing the development of the technical basis to demonstrate the feasibility of storing UNF for extended periods of time with subsequent transportation of the UNF to its final disposition. This program is fully integrated with industry, the U.S. regulator, and the international community to assure that programmatic goals and objectives are consistent with a broad perspective of technical and regulatory opinion. As the work evolves, assessments will be made to ensure that the work continues to focus on the overall goals and objectives of the program. (authors)

  3. Test report for PAS-1 cask certification for shipping payload B

    SciTech Connect (OSTI)

    MERCADO, J.E.

    1998-10-13T23:59:59.000Z

    This test report documents the successful inspection and testing to certify two NuPac PAS-1 casks in accordance with US Department of Energy Certificate of Compliance (CoC) USA/9184/B(U). The primary and secondary containment vessels of each cask met the acceptance criteria defined in the CoC and the test plan.

  4. Drying Fruits and Vegetables at Home.

    E-Print Network [OSTI]

    Putnam, Peggy H.

    1981-01-01T23:59:59.000Z

    that are responsible for their maturation, or their becoming ripe. These enzymes cause color and flavor changes, some of which may become more extensive when food surfaces are cut and exposed to air. The changes con tinue during drying and storage unless the enzyme... in recommendations for treatment before dry ing, for methods of drying, for temperatures and length of drying time, and for conditioning prior to storage. You may have to use the "trial and error" approach in finding out which drying technique works best for your...

  5. Well injectivity during CO2 storage operations in deep saline aquifers6 1: Experimental investigation of drying effects, salt precipitation and7

    E-Print Network [OSTI]

    Boyer, Edmond

    Carbon Capture and Storage (CCS) is a technique than can potentially limit the accumulation29-17Jan2014 #12;3 1. Introduction51 52 Geological sequestration of CO2 into deep saline aquifers studied54 much less than mature oil & gas reservoirs. Injection of carbon dioxide into saline aquifers55

  6. Well injectivity during CO2storage operations in deep saline aquifers 6 Part 2: Numerical simulations of drying, salt deposit mechanisms and role of7

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 1 2 3 4 5 Well injectivity during CO2storage operations in deep saline aquifers ­6 Part 2 or industrial units and stored in underground geological reservoirs.30 Return on experience withCO2 injection-well field scale is proposed. This approach is of major40 importance because it makes it possible

  7. A source-term method for determining spent-fuel transport cask containment requirements: Executive summary

    SciTech Connect (OSTI)

    Sanders, T.L.; Seager, K.D. (Sandia National Labs., Albuquerque, NM (United States)); Reardon, P.C. (GRAM, Inc., Albuquerque, NM (United States))

    1993-02-01T23:59:59.000Z

    This Executive Summary presents the methodology for determining containment requirements for spent-fuel transport casks under normal and hypothetical accident conditions. Three sources of radioactive material are considered: (1) the spent fuel itself, (2) radioactive material, referred to as CRUD, attached to the outside surfaces of fuel rod cladding, and (3) residual contamination adhering to interior surfaces of the cask cavity. The methodologies for determining the concentrations of freely suspended radioactive materials within a spent-fuel transport cask for these sources are discussed in much greater detail in three companion reports: A Method for Determining the Spent-Fuel Contribution to Transport Cask Containment Requirements,'' Estimate of CRUD Contribution to Shipping Cask Containment Requirements,'' and A Methodology for Estimating the Residual Contamination Contribution to the Source Term in a Spent-Fuel Transport Cask.'' Examples of cask containment requirements that combine the individually determined containment requirements for the three sources are provided, and conclusions from the three companion reports to this Executive Summary are presented.

  8. ANSI N14.5 source term licensing of spent-fuel transport cask containment

    SciTech Connect (OSTI)

    Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States); James, R.J.; Foadian, H.; Rashid, Y.R. [ANATECH Research Corp., La Jolla, CA (United States)

    1993-10-01T23:59:59.000Z

    American National Standards Institute (ANSI) standard N14.5 states that ``compliance with package containment requirements shall be demonstrated either by determination of the radioactive contents release rate or by measurement of a tracer material leakage rate.`` The maximum permissible leakage rate from the transport cask is equal to the maximum permissible release rate divided by the time-averaged volumetric concentration of suspended radioactivity within the cask. The development of source term methodologies at Sandia National Laboratories (SNL) provides a means to determine the releasable radionuclide concentrations within spent-fuel transport casks by estimating the probability of cladding breach, quantifying the amount of radioactive material released into the cask interior from the breached fuel rods, and quantifying the amount of radioactive material within the cask due to other sources. These methodologies are implemented in the Source Term Analyses for Containment Evaluations (STACE) software. In this paper, the maximum permissible leakage rates for the normal and hypothetical accident transport conditions defined by 10 CFR 71 are estimated using STACE for a given cask design, fuel assembly, and initial conditions. These calculations are based on defensible analysis techniques that credit multiple release barriers, including the cladding and the internal cask walls.

  9. Pilot study dismantlement of 20 lead-lined shipping casks

    SciTech Connect (OSTI)

    Thurmond, S.M.

    1995-08-01T23:59:59.000Z

    This report describes a pilot study conducted at the INEL to dismantle lead-lined casks and shielding devices, separate the radiologically contaminated and hazardous materials, and recycle resultant scrap lead. The facility areas where the work was performed, dismantlement methods, and process equipment are described. Issues and results associated with recycling the lead as a free-released scrap metal are presented and discussed. Data and results from the pilot study are summarized and presented. The study concluded that cask dismantlement at the INEL can be performed as a legitimate recycling activity for scrap lead. Ninety-one percent of the lead recovered passed free-release criteria. The value of the 50,375 lb of recovered lead is approximately $0.45/lb. Resultant waste streams can be satisfactorily treated and disposed. Only very low levels of bulk radiological contamination (47 picocuries/gram of 137 Cs and 3.2 picocuries/gram of {sup 6O}Co) were detected in the lead rejected for free release.

  10. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    SciTech Connect (OSTI)

    Not Available

    1980-05-01T23:59:59.000Z

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  11. A method for determining the spent-fuel contribution to transport cask containment requirements

    SciTech Connect (OSTI)

    Sanders, T.L.; Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R.; Barrett, P.R. [ANATECH Research Corp., La Jolla, CA (United States); Malinauskas, A.P. [Oak Ridge National Lab., TN (United States); Einziger, R.E. [Pacific Northwest Lab., Richland, WA (United States); Jordan, H. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant; Duffey, T.A.; Sutherland, S.H. [APTEK, Inc., Colorado Springs, CO (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States)

    1992-11-01T23:59:59.000Z

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  12. Documentation for first annual testing and inspections of Benificial Uses Shipping System (BUSS) Cask

    SciTech Connect (OSTI)

    Lundeen, J.E.

    1994-08-23T23:59:59.000Z

    The purpose of this report is to compile date generated during the first annual tests and inspections of the Benificiai Uses Shipping System (BUSS) Cask. In addition, this report will verify that the testing criteria identified in chapter 8 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met. Section 8.2 ``Maintenance and Periodic Inspection Program`` of the BUSS Cask SARP requires that the following tests and inspections be performed on an annual basis: Hydrostatic pressure test; helium leak test; dye penetrant test on the trunnions and lifting lugs; and torque test on all bolts; impact limiter inspection and weight test. The first annual inspections and testing of the BUSS Cask were completed on May 5, 1994, and met the SARP criteria.

  13. CRITICALITY SAFETY CONTROL OF LEGACY FUEL FOUND AT 105-K WEST FUEL STORAGE BASIN

    SciTech Connect (OSTI)

    JENSEN, M.A.

    2005-08-19T23:59:59.000Z

    In August 2004, two sealed canisters containing spent nuclear fuel were opened for processing at the Hanford Site's K West fuel storage basin. The fuel was to be processed through cleaning and sorting stations, repackaged into special baskets, placed into a cask, and removed from the basin for further processing and eventual dry storage. The canisters were expected to contain fuel from the old Hanford C Reactor, a graphite-moderated reactor fueled by very low-enriched uranium metal. The expected fuel type was an aluminum-clad slug about eight inches in length and with a weight of about eight pounds. Instead of the expected fuel, the two canisters contained several pieces of thin tubes, some with wire wraps. The material was placed into unsealed canisters for storage and to await further evaluation. Videotapes and still photographs of the items were examined in consultation with available retired Hanford employees. It was determined that the items had a fair probability of being cut-up pieces of fuel rods from the retired Hanford Plutonium Recycle Test Reactor (PRTR). Because the items had been safely handled several times, it was apparent that a criticality safety hazard did not exist when handling the material by itself, but it was necessary to determine if a hazard existed when combining the material with other known types of spent nuclear fuel. Because the PRTR operated more than 40 years ago, investigators had to rely on a combination of researching archived documents, and utilizing common-sense estimates coupled with bounding assumptions, to determine that the fuel items could be handled safely with other spent nuclear fuel in the storage basin. As older DOE facilities across the nation are shut down and cleaned out, the potential for more discoveries of this nature is increasing. As in this case, it is likely that only incomplete records will exist and that it will be increasingly difficult to immediately characterize the nature of the suspect fissionable material and its criticality hazards.

  14. Spent Fuel Transportation Cask Response to the Caldecott Tunnel Fire Scenario

    SciTech Connect (OSTI)

    Adkins, Harold E.; Koeppel, Brian J.; Cuta, Judith M.

    2007-01-01T23:59:59.000Z

    On April 7, 1982, a tank truck and trailer carrying 8,800 gallons of gasoline was involved in an accident in the Caldecott tunnel on State Route 24 near Oakland, California. The tank trailer overturned and subsequently caught fire. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook analyses to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by truck. The Fire Dynamics Simulator (FDS) code developed by National Institute of Standards and Technology (NIST) was used to determine the thermal environment in the Caldecott tunnel during the fire. The FDS results were used to define boundary conditions for a thermal transient model of a truck transport cask containing spent nuclear fuel. The Nuclear Assurance Corporation (NAC) Legal Weight Truck (LWT) transportation cask was selected for this evaluation, as it represents a typical truck (over-the-road) cask, and can be used to transport a wide variety of spent nuclear fuels. Detailed analysis of the cask response to the fire was performed using the ANSYS® computer code to evaluate the thermal performance of the cask design in this fire scenario. This report describes the methods and approach used to assess the thermal response of the selected cask design to the conditions predicted in the Caldecott tunnel fire. The results of the analysis are presented in detail, with an evaluation of the cask response to the fire. The staff concluded that some components of smaller transportation casks resembling the NAC LWT, despite placement within an ISO container, could degrade significantly. Small transportation casks similar to the NAC LWT would probably experience failure of seals in this severe accident scenario. USNRC staff evaluated the radiological consequences of the cask response to the Caldecott tunnel fire. Although some components heated up beyond their service temperatures, the staff determined that there would be no significant release as a result of the fire for the NAC LWT and similar casks.

  15. DRY TRANSFER FACILITY WORKER DOSE ASSESSMENT

    SciTech Connect (OSTI)

    J.S. Tang

    2004-09-23T23:59:59.000Z

    The purpose of this calculation is to estimate radiation doses received by personnel working in the Dry Transfer Facility No.1 (DTF-1) performing operations to receive transportation casks, transfer wastes, prepare waste packages, and ship out loaded waste packages and empty casks. Doses received by workers due to maintenance operations are also included in this revision. The specific scope of work contained in this calculation covers both collective doses and individual worker group doses on an annual basis, and includes the contributions due to external and internal radiation from normal operation, excluding the remediation area of the building. The results of this calculation will be used to support the design of the DTF-1 and to provide occupational dose estimates for the License Application. The calculations contained in this document were developed by Environmental and Nuclear Engineering of the Design and Engineering Organization and are intended solely for the use of the Design and Engineering Organization in its work regarding facility operation. Yucca Mountain Project personnel from the Environmental and Nuclear Engineering should be consulted before use of the calculations for purposes other than those stated herein or use by individuals other than authorized personnel in the Environmental and Nuclear Engineering.

  16. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    SciTech Connect (OSTI)

    none,

    1990-02-01T23:59:59.000Z

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed.

  17. Operations manual for the Beneficial Uses Shipping System cask. Revision 1

    SciTech Connect (OSTI)

    Bronowski, D.R.; Yoshimura, H.R.

    1993-04-01T23:59:59.000Z

    This document is the Operations Manual for the Beneficial Uses Shipping System (BUSS) cask. These operating instructions address requirements; for loading, shipping, and unloading, supplementing general operational information found in the BUSS Safety Analysis Report for Packaging (SARP), SAND 83-0698. Use of the BUSS cask is authorized by Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) for the shipment of special form cesium chloride or strontium flouride capsules.

  18. Maintenance manual for the Beneficial Uses Shipping System cask. Revision 1

    SciTech Connect (OSTI)

    Bronowski, D.R.; Yoshimura, H.R.

    1993-05-01T23:59:59.000Z

    This document is the Maintenance Manual for the Beneficial Uses Shipping System (BUSS) cask. These instructions address requirements for maintenance, inspection, testing, and repair, supplementing general information found in the BUSS Safety Analysis Report for Packaging (SARP), SAND 83-0698. Use of the BUSS cask is authorized by the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) for the shipment of special form cesium chloride or strontium flouride capsules.

  19. Environmental Assessment: Relocation and storage of TRIGA{reg_sign} reactor fuel, Hanford Site, Richland, Washington

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    In order to allow the shutdown of the Hanford 308 Building in the 300 Area, it is proposed to relocate fuel assemblies (101 irradiated, three unirradiated) from the Mark I TRIGA Reactor storage pool. The irradiated fuel assemblies would be stored in casks in the Interim Storage Area in the Hanford 400 Area; the three unirradiated ones would be transferred to another TRIGA reactor. The relocation is not expected to change the offsite exposure from all Hanford Site 300 and 400 Area operations.

  20. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil)] [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil)] [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina)] [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)] [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01T23:59:59.000Z

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  1. Spent fuel drying system test results (second dry-run)

    SciTech Connect (OSTI)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01T23:59:59.000Z

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks have been detected in the basins and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the second dry-run test, which was conducted without a fuel element. With the concurrence of project management, the test protocol for this run, and subsequent drying test runs, was modified. These modifications were made to allow for improved data correlation with drying procedures proposed under the IPS. Details of these modifications are discussed in Section 3.0.

  2. Spent-fuel-storage alternatives

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  3. STP-ECRTS - THERMAL AND GAS ANALYSES FOR SLUDGE TRANSPORT AND STORAGE CONTAINER (STSC) STORAGE AT T PLANT

    SciTech Connect (OSTI)

    CROWE RD; APTHORPE R; LEE SJ; PLYS MG

    2010-04-29T23:59:59.000Z

    The Sludge Treatment Project (STP) is responsible for the disposition of sludge contained in the six engineered containers and Settler tank within the 105-K West (KW) Basin. The STP is retrieving and transferring sludge from the Settler tank into engineered container SCS-CON-230. Then, the STP will retrieve and transfer sludge from the six engineered containers in the KW Basin directly into a Sludge Transport and Storage Containers (STSC) contained in a Sludge Transport System (STS) cask. The STSC/STS cask will be transported to T Plant for interim storage of the STSC. The STS cask will be loaded with an empty STSC and returned to the KW Basin for loading of additional sludge for transportation and interim storage at T Plant. CH2MHILL Plateau Remediation Company (CHPRC) contracted with Fauske & Associates, LLC (FAI) to perform thermal and gas generation analyses for interim storage of STP sludge in the Sludge Transport and Storage Container (STSCs) at T Plant. The sludge types considered are settler sludge and sludge originating from the floor of the KW Basin and stored in containers 210 and 220, which are bounding compositions. The conditions specified by CHPRC for analysis are provided in Section 5. The FAI report (FAI/10-83, Thermal and Gas Analyses for a Sludge Transport and Storage Container (STSC) at T Plant) (refer to Attachment 1) documents the analyses. The process considered was passive, interim storage of sludge in various cells at T Plant. The FATE{trademark} code is used for the calculation. The results are shown in terms of the peak sludge temperature and hydrogen concentrations in the STSC and the T Plant cell. In particular, the concerns addressed were the thermal stability of the sludge and the potential for flammable gas mixtures. This work was performed with preliminary design information and a preliminary software configuration.

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

    SciTech Connect (OSTI)

    LANGEVIN, A.S.

    1999-07-12T23:59:59.000Z

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

  5. Safety analysis report vitrified high level waste type B shipping cask

    SciTech Connect (OSTI)

    NONE

    1995-03-01T23:59:59.000Z

    This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

  6. MicroShield analysis to calculate external radiation dose rates for several spent fuel casks

    SciTech Connect (OSTI)

    Marincel, M.K. [Missouri Univ., Rolla, MO (United States); Weiner, R.F.; Osborn, D.M. [Sandia National Laboratories, Albuquerque, NM (United States)

    2007-07-01T23:59:59.000Z

    The purpose of this MicroShield analysis is to calculate the external radiation, primarily gamma, dose rate for spent fuel casks. The reason for making this calculation is that currently all analyses of transportation risk assume that this external dose rate is the maximum allowed by regulation, 10 mrem/hr at 2 m from the casks, and the risks of incident-free transportation are thus always overestimated to an unknown extent. In order to do this, the program by Grove Software, MicroShield 7.01, was used to model three Nuclear Regulatory Commission (NRC) approved casks: HI-STAR 100, GA-4, and NAC-STC, loaded with specific source material. Dimensions were obtained from NUREG/CR-6672 and the Certificates of Compliance for each respective cask. Detectors were placed at the axial point at 1 m and 2 m from the outer gamma shielding of the casks. In the April 8, 2004 publication of the Federal Register, a notice of intent to prepare a Supplemental Yucca Mountain Environmental Impact Statement (DOE/EIS-0250F-S1) was published by the Office of Civilian Radioactive Waste Management (OCRWM) in order to consider design, construction, operation, and transportation of spent nuclear fuel to the Yucca Mountain repository [1]. These more accurate estimates of the external dose rates could be used in order to provide a more risk-informed analysis. (authors)

  7. Annotated Bibliography for Drying Nuclear Fuel

    SciTech Connect (OSTI)

    Rebecca E. Smith

    2011-09-01T23:59:59.000Z

    Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

  8. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    SciTech Connect (OSTI)

    Sanders, T.L. (Sandia National Labs., Albuquerque, NM (United States)); Jordan, H. (EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant); Pasupathi, V. (Battelle, Columbus, OH (United States)); Mings, W.J. (USDOE Idaho Field Office, Idaho Falls, ID (United States)); Reardon, P.C. (GRAM, Inc., Albuquerque, NM (United States))

    1991-09-01T23:59:59.000Z

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs.

  9. Impacts of SNF burnup credit on the shipment capability of the GA-4 cask

    SciTech Connect (OSTI)

    Mobasheran, A.S. [Roy F. Weston, Inc., Washington, DC (United States); Lake, W. [Department of Energy, Washington, DC (United States); Richardson, J. [Raytheon Nuclear Inc., Washington, DC (United States)

    1996-12-01T23:59:59.000Z

    Scoping analyses were performed to determine the impacts of two different levels of burnup credit and two different spent fuel pickup rates on the shipment capability and the minimum fleet size of the GA-4 cask. The analyses involved developing loading curves for the GA-4 cask based on the actinide-only and principal-isotope burnup credit considerations. The analyses also involved examination of the spent nuclear fuel assembly population at nine reactor sites and categorization of the assemblies in accordance with the loading restrictions imposed. The results revealed that for the nine sites considered, depending on the level of burnup credit and the pickup rate assumed, the total savings in shipment and cask fleet costs (1994 dollars) can range from $55 million to $74 million.

  10. A full-scale thermal test and analytical evaluation of the beneficial uses shipping system cask

    SciTech Connect (OSTI)

    Moya, J.L.; Akau, R.L.

    1988-09-01T23:59:59.000Z

    A thermal test of the Beneficial Uses Shipping System (BUSS) cask containing irradiation source capsules was conducted to verify a two-dimensional axisymmetric thermal model developed for the Safety Analysis Report. The BUSS cask is a Type B package developed to transport irradiation source capsules of cesium chloride or strontium fluoride to commercially licensed food and pharmaceutical irradiating facilities. The uniqueness of this test is that it was performed on an internally instrumented, full-scale cask with actual radioactive capsules. This resulted in more realistic system temperatures than those obtained if heaters were used to simulate the large gamma source. In addition, the thermal test provides benchmark data for other thermal codes. 12 refs.; 24 figs.; 2 tabs.

  11. TMI-2 (Three-Mile Island-Unit 2) rail cask and railcar maintenance

    SciTech Connect (OSTI)

    Tyacke, M.J.; Ayers, A.L., Jr.; Ball, L.J.; Anselmo, A.A.

    1988-02-01T23:59:59.000Z

    This paper describes the NuPac 125-B cask system (i.e., cask and railcar), and the maintenance and inspection requirements for that system. The paper discusses the operations being done to satisfy those requirements and how, in some cases, it has been efficient for the operations to be more rigorous than the requirements. Finally, this paper discusses the experiences gained from those operations and how specific hardware and procedural enhancements have resulted in a reliable and continuous shipping campaign. 2 refs., 2 figs.

  12. Review of Drying Methods for Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Large, W.S.

    1999-10-21T23:59:59.000Z

    SRTC is developing technology for direct disposal of aluminum spent nuclear fuel (SNF). The development program includes analyses and tests to support design and safe operation of a facility for ''road ready'' dry storage of SNF-filled canisters. The current technology development plan includes review of available SNF drying methods and recommendation of a drying method for aluminum SNF.

  13. Storage stability of human milk enzymes

    E-Print Network [OSTI]

    Chen, Chia-Tsun

    1980-01-01T23:59:59.000Z

    . Those sub- stances are important to the health of infants and pre- mature babies. Room temperature (25 0) storage of freeze-dried m'lk, low temperature (-20 C) storage of freeze-dried milk, and low temperature (-20 0) stcrage o " liouid milk were... selected to meet the need of esta'blis'n- ing the human milk bank. Lipase activity decreased sharply during the first day of' storage at 20 0 for freeze-dried samples, and at -20~0 f' or freeze-dried samples as well as in those milks f'rozen from...

  14. CASK/MSC/WP PREPARATION SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    S. Drummond

    2005-04-12T23:59:59.000Z

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the Cask/MSC/WP preparation system and their bases to allow the design effort to proceed to license application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design, as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This type of SDD both leads and trails the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. This SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. This SDD addresses the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]) requirements. Additional PRD requirements may be cited, as applicable, to drive the design of specific aspects of the system, with justifications provided in the basis. Functional and operational requirements applicable to this system are obtained from the ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]) document. Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (BSC 2004 [DIRS 171599]) and the preclosure safety analyses.

  15. Drying Foods at Home Safely Drying Herbs

    E-Print Network [OSTI]

    jars, freezer bags, and airtight plastic containers. Like other foods dried at home, dried herbs in an airtight container and store in a cool, dry, and dark place. Recommended containers include glass canning

  16. Dry effluent

    SciTech Connect (OSTI)

    Brady, J.D. (Anderson, 2000 Inc., Peachtree City, GA (US))

    1988-01-01T23:59:59.000Z

    The available choices of pollution control systems depend on what is being burned and how stringent the regulations are. The common systems are gas cooling by a waste heat boiler or an air-air heat exchanger followed by fabric filtration or electrostatic precipitation for particulate removal; alkaline spray absorbers followed by fabric filters (dry scrubbers) for particulate and acid gas removal; wet scrubbers for simultaneous particulate and acid gas removal, and; the newest - spray evaporation, followed by wet scrubbing for particulate and acid gas removal. Each has advantages and each has disadvantages. This paper discusses the advantages and disadvantages of the spray evaporator and wet scrubber combination.

  17. Geometric Feasibility of ITER Air Cushion Remote Handling Casks and Extensions for Free

    E-Print Network [OSTI]

    Ribeiro,Isabel

    -NET/97-460 (EFDA) | 1998 From lift to laydown hall Path: Cubic Spiral wheels following different paths to fully explore the rhombic configuration From lift to VV gallery Path: Cubic Spiral #12;Geometric Cask artificial landmarks Laser scanner steering sensor guiding tape deformed magnetic field Automated

  18. Drying and Storing Sorghum Grain.

    E-Print Network [OSTI]

    Hutchison, J. E.

    1959-01-01T23:59:59.000Z

    Drying and Storing Sorghum Grain W. S. ALLEN AND J. W. SORENSON. JR.* lead to insect. niold and heat damage in stored grain. They cause most of the problems encountered in storing grain. High moisture may result from leak- age of outside... moisture through hin walls or from placing high-moisture grain in storage. If the following recornrnendations and procedures are followed. sorghum grain can be stored safely. The! are based on research conducted at Beeville by the Texas Agricultural...

  19. Second Annual Maintenance, Inspection, and Test Report for PAS-1 Cask Certification for Shipping Payload B

    SciTech Connect (OSTI)

    KELLY, D.J.

    2000-10-09T23:59:59.000Z

    The Nuclear Packaging, Inc. (NuPac), PAS-1 cask is required to undergo annual maintenance and inspections to retain certification in accordance with U.S. Department of Energy (DOE) Certificate of Compliance USA/9184B(U) (Appendix A). The packaging configuration being tested and maintained is the NuPac PAS-1 cask for Payload B. The intent of the maintenance and inspections is to ensure the packaging remains in unimpaired physical condition. Two casks, serial numbers 2162-026 and 2162-027, were maintained, inspected, and tested at the 306E Development, Fabrication, and Test Laboratory, located at the Hanford Site's 300 Area. Waste Management Federal Services, Inc. (WMFS), a subsidiary of GTS Duratek, was in charge of the maintenance and testing. Cogema Engineering Corporation (Cogema) directed the operations in the test facility. The maintenance, testing, and inspections were conducted successfully with both PAS-1 casks. The work conducted on the overpacks included weighing, gasket replacement, and plastic pipe plug and foam inspections. The work conducted on the secondary containment vessel (SCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, and a helium leak test. The work conducted on the primary containment vessel (PCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, dimensional inspection of the vessel bottom, a helium leak test, and dye penetrant inspection of the welds. The vermiculite material used in the cask rack assembly was replaced.

  20. Operations of the LR56 radioactive liquid cask transport system at U.S. Department of Energy sites

    SciTech Connect (OSTI)

    Davidson, J.S. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States); Hornstra, D.J. [Performance Development Corp., Oak Ridge, TN (United States); Sazawal, V.K. [NUMATEC, Inc., Bethesda, MD (United States); Clement, G. [SGN, St. Quentin en Yvelines (France)

    1996-06-01T23:59:59.000Z

    The LR56 cask system is licensed for use in France under Certificate of Compliance F/309/B(U)F for transport of 4,000-liter volumes of radioactive liquids. Three LR56 cask systems (with modifications for use at Department of Energy (DOE) sites) have been purchased for delivery at the Hanford Site, Oak Ridge National Laboratory (ORNL), and Savannah River Site (SRS). The LR56 cask systems will be used for on-site transfers of Type B quantities of radioactive liquid waste. The ORNL unit will also be used as a Type A packaging for transfers of radioactive liquids between DOE sites. This paper discusses LR56 operating features and the use of the cask system at the three DOE sites.

  1. Preliminary analysis of the postulated changes needed to achieve rail cask handling capabilities at selected light water reactors

    SciTech Connect (OSTI)

    Konzek, G.J.

    1986-02-01T23:59:59.000Z

    Reactor-specific railroad and crane information for all LWRs in the US was extracted from current sources of information. Based on this information, reactors were separated into two basic groups consisting of reactors with existing, usable rail cask capabilities and those without these capabilities. The latter group is the main focus of this study. The group of reactors without present rail cask handling capabilities was further separated into two subgroups consisting of reactors considered essentially incapable of handling a large rail cask of about 100 tons and reactors where postulated facility changes could result in rail cask handling capabilities. Based on a selected population of 127 reactors, the results of this assessment indicate that usable rail cask capabilities exist at 83 (65%) of the reactors. Twelve (27%) of the remaining 44 reactors are deemed incapable of handling a large rail cask without major changes, and 32 reactors are considered likely candidates for potentially achieving rail cask handling capabilities. In the latter group, facility changes were postulated that would conceptually enable these reactors to handle large rail casks. The estimated cost per plant of required facility changes varied widely from a high of about $35 million to a low of <$0.3 million. Only 11 of the 32 plants would require crane upgrades. Spur track and right-of-way costs would apparently vary widely among sites. These results are based on preliminary analyses using available generic cost data. They represent lower bound values that are useful for developing an initial assessment of the viability of the postulated changes on a system-wide basis, but are not intended to be absolute values for specific reactors or sites.

  2. APS Protocols for Handling, Storage, and Disposal of Untreated...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Heat-treat wooden, metal, or cardboard shipping containers (using lowest heat). Treat plastic containers and coolers with 70% ethyl alcohol. Storage of Samples: Store dry samples...

  3. Spent fuel drying system test results (first dry-run)

    SciTech Connect (OSTI)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01T23:59:59.000Z

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first dry-run test, which was conducted without a fuel element. The empty test apparatus was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The data from this dry-run test can serve as a baseline for the first two fuel element tests, 1990 (Run 1) and 3128W (Run 2). The purpose of this dry-run was to establish the background levels of hydrogen in the system, and the hydrogen generation and release characteristics attributable to the test system without a fuel element present. This test also serves to establish the background levels of water in the system and the water release characteristics. The system used for the drying test series was the Whole Element Furnace Testing System, described in Section 2.0, which is located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in section 3.0, and the experimental results provided in Section 4.0. These results are further discussed in Section 5.0.

  4. WIPP Remote Handled Waste Facility: Performance Dry Run Operations

    SciTech Connect (OSTI)

    Burrington, T. P.; Britain, R. M.; Cassingham, S. T.

    2003-02-24T23:59:59.000Z

    The Remote Handled (RH) TRU Waste Handling Facility at the Waste Isolation Pilot Plant (WIPP) was recently upgraded and modified in preparation for handling and disposal of RH Transuranic (TRU) waste. This modification will allow processing of RH-TRU waste arriving at the WIPP site in two different types of shielded road casks, the RH-TRU 72B and the CNS 10-160B. Washington TRU Solutions (WTS), the WIPP Management and Operation Contractor (MOC), conducted a performance dry run (PDR), beginning August 19, 2002 and successfully completed it on August 24, 2002. The PDR demonstrated that the RHTRU waste handling system works as designed and demonstrated the handling process for each cask, including underground disposal. The purpose of the PDR was to develop and implement a plan that would define in general terms how the WIPP RH-TRU waste handling process would be conducted and evaluated. The PDR demonstrated WIPP operations and support activities required to dispose of RH-TRU waste in the WIPP underground.

  5. Preconceptual design for a Monitored Retrievable Storage (MRS) transfer facility

    SciTech Connect (OSTI)

    Woods, W.D.; Jowdy, A.K. (Parsons (Ralph M.) Co., Pasadena, CA (USA)); Smith, R.I. (Pacific Northwest Lab., Richland, WA (USA))

    1990-09-01T23:59:59.000Z

    The contract between the DOE and the utilities specifies that the DOE will receive spent fuel from the nuclear utilities in 1998. This study investigates the feasibility of employing a simple Transfer Facility which can be constructed quickly, and operate while the full-scale MRS facilities are being constructed. The Transfer Facility is a hot cell designed only for the purpose of transferring spent fuel assemblies from the Office of Civilian Radioactive Waste Management (OCRWM) transport casks (shipped from the utility sites) into onsite concrete storage casks. No operational functions other than spent fuel assembly transfers and the associated cask handling, opening, and closing would be performed in this facility. Radioactive waste collected in the Transfer Facility during operations would be stored until the treatment facilities in the full-scale MRS facility became operational, approximately 2 years after the Transfer Facility started operation. An alternate wherein the Transfer Facility was the only waste handling building on the MRS site was also examined and evaluated. 6 figs., 26 tabs.

  6. Factors affecting the recovery of bacteria in freeze-dried model systems

    E-Print Network [OSTI]

    Custer, Carl Steven

    1970-01-01T23:59:59.000Z

    cellulose, and subsequently freeze-dried. The influences of the rate of freezing, time and atmosphere of storage, position of the bacteria within the model system and condition of incubation upon the survival and metabolic injury of the freeze-dried... storage. The type of atmosphere in which freeze-dried preparations were stored was important to bacterial survival. Nitrogen was less damaging to freeze-dried cells than air. Strata studies indicated that the viable bacterial population density...

  7. The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

    SciTech Connect (OSTI)

    Phillips, C.; Thomas, I. [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA 99354 (United States)

    2013-07-01T23:59:59.000Z

    This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility, the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.

  8. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    SciTech Connect (OSTI)

    Romano, T.

    1997-09-29T23:59:59.000Z

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  9. Incentives for the use of depleted uranium alloys as transport cask containment structure

    SciTech Connect (OSTI)

    McConnell, P [GRAM, Inc., Albuquerque, NM (United States); Salzbrenner, R; Wellman, G W; Sorenson, K B [Sandia National Labs., Albuquerque, NM (United States)

    1992-01-01T23:59:59.000Z

    Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both nominal use and accident conditions to serve the dual-role of shielding and containment, the use of other structure materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describesa two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature. The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracmm resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as win be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term structural credit'' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.)

  10. Final design review summary report for the TN-WHC cask and transportation system

    SciTech Connect (OSTI)

    Kee, A.T.

    1997-01-17T23:59:59.000Z

    This document represents comments generated from a review of Transnuclear`s Final Design Package distributed on December 10, 1996 and a review of the Final Design Analysis Report meeting held on December 17 & 18, 1996. The Final design describes desicn features and presents final analyses @j performed to fabricate and operate the system while meeting the Cask/Transportation Functions and Requirements, WHC-SD-SNF-FRD-011, Rev. 0 and specification WHC-S-0396, Rev. 1.

  11. Nuclear Criticality Safety Evaluation of the 9965, 9968, 9972, 9973, 9974, and 9975 Shipping Casks

    SciTech Connect (OSTI)

    Frost, R.L.

    1999-02-26T23:59:59.000Z

    A Nuclear Criticality Safety Evaluation (NCSE) has been performed for the 9965, 9968, 9972, 9973, 9974, and 9975 SRS-designed shipping casks. This was done in support of the recertification effort for the 9965 and 9968, and the certification of the newly designed 9972-9975 series. The analysis supports the use of these packages as Fissile Class I for shipment of fissionable material from the SRS FB-Line, HB-Line, and from Lawrence Livermore national Laboratory. six different types of material were analyzed with varying Isotopic composition, of both oxide and metallic form. The mass limits required to support the fissile Class I rating for each of the envelopes are given in the Table below. These mass limits apply if DOE approves an exception as described in 10 CFR 71.55(c), such that water leakage into the primary containment vessel does not need to be considered in the criticality analysis. If this exception is not granted, the mass limits are lower than those shown below. this issue is discussed in detail in sections 5 and 6 of the report.One finding from this work is important enough to highlight in the abstract. The fire tests performed for this family of shipping casks indicates only minimal charring of the Celotex thermal insulation. Analysis of the casks with no Celotex insulation (assuming it has all burned away), results in values of k-eff that exceed 1.0. Therefore, the Celotex insulation must remain intact in order to guarantee sub criticality of the 9972-9975 family of shipping casks.

  12. Criticality assessment of basket designs for use in the MH-1A shipping cask

    SciTech Connect (OSTI)

    Thomas, J.T.

    1982-04-01T23:59:59.000Z

    An analytical study is made of a proposed stainless steel basket to be used in the MH-1A cask for the shipment of National Bureau of Standards reactor fuel elements. The use of B/sub 4/C in the basket as a primary control for the prevention of criticality is shown to be effective but not necessary. A basket fabricated of 0.635-cm-thick stainless steel provides a sufficient margin of subcriticality.

  13. QA in the design and fabrication of the TMI-2 rail cask

    SciTech Connect (OSTI)

    Hayes, G.R.

    1988-01-01T23:59:59.000Z

    EGandG Idaho, Inc., acting on behalf of the US Department of Energy, is responsible for transporting core debris from Three Mile Island-Unit 2 to the Idaho National Engineering Laboratory. Transportation of the debris is being accomplished using an NRC licensed container, called the NuPac 125-B. This paper describes the NuPac 125-B Rail Cask and the quality assurance (QA) requirements for that system. Also discussed are the QA roles of the various organizations involved in designing, building, inspecting and testing the NuPac 125-B. The paper presents QA/QC systems implemented during the design, procurement, and fabrication of the cask to assure compliance with all applicable technical codes, standards and regulations. It also goes beyond the requirements aspect and describes unique QA/QC measures employed to assure that the cask was built with minimum QA problems. Finally, the lessons learned from the NuPac 125-B project is discussed. 4 refs., 4 figs.

  14. STRUCTURAL ANALYSES OF FUEL CASKS SUBJECTED TO BOLT PRELOAD, INTERNAL PRESSURE AND SEQUENTIAL DYNAMIC IMPACTS

    SciTech Connect (OSTI)

    Wu, T

    2009-06-25T23:59:59.000Z

    Large fuel casks subjected to the combined loads of closure bolt tightening, internal pressure and sequential dynamic impacts present challenges when evaluating their performance in the Hypothetical Accident Conditions (HAC) specified in the Code of Federal Regulations Title 10 Part 71 (10CFR71). Testing is often limited by cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing by using simplified analytical methods. In addition, there are no realistic analyses of closure bolt stresses for HAC conditions reported in the open literature. This paper presents a numerical technique for analyzing the accumulated damages of a large fuel cask caused by the sequential loads of the closure bolt tightening and the internal pressure as well as the drop and crash dynamic loads. The bolt preload and the internal pressure are treated as quasi-static loads so that the finite element method with explicit numerical integration scheme based on the theory of wave propagation can be applied. The dynamic impacts with short durations such as the 30-foot drop and the 40-inch puncture for the hypothetical accident conditions specified in 10CFR71 are also analyzed by using the finite-element method with explicit numerical integration scheme.

  15. Cold Vacuum Drying (CVD) Facility General Service Helium System Design Description

    SciTech Connect (OSTI)

    FARWICK, C.C.

    1999-07-06T23:59:59.000Z

    This document describes the Cold Vacuum Drying Facility general service helium system (GSHe). The GSHe is a general service facility process support system, but does include safety-class systems, structures and components providing protection to the offsite public. The GSHe also performs safety-significant functions that provide protection to onsite workers. The GSHe essential function is to provide helium to support process functions during all phases of facility operations. GSHe helium is used to purge the cask and the MCO in order to maintain their internal atmospheres below hydrogen flammability concentrations. The GSHe also supplies helium to purge the PWC lines and components and the VPS vacuum pump.

  16. Project W-441, cold vacuum drying facility design requirements document

    SciTech Connect (OSTI)

    O`Neill, C.T.

    1997-05-08T23:59:59.000Z

    This document has been prepared and is being released for Project W-441 to record the design basis for the design of the Cold Vacuum Drying Facility. This document sets forth the physical design criteria, Codes and Standards, and functional requirements that were used in the design of the Cold Vacuum Drying Facility. This document contains section 3, 4, 6, and 9 of the Cold Vacuum Drying Facility Design Requirements Document. The remaining sections will be issued at a later date. The purpose of the Facility is to dry, weld, and inspect the Multi-Canister Overpacks before transport to dry storage.

  17. Biomass Logistics and Particle Technology Group Purdue Improved Drying

    E-Print Network [OSTI]

    Ginzel, Matthew

    to maintain quality of grain in storage. n Farmers primarily depended on open air solar drying after logistics Grain & pest management Pre-Harvest losses from: Insect, molds and birds Harvesting & handling of PICS, technology Open Air Solar Drying of Maize in Ejura Market, Ashanti Region, Ghana #12;4 Chronology

  18. Shielding analysis for the 300 area light water reactor spent nuclear fuel within a modified multi-canister overpack canister in a modified multi-canister overpack cask

    SciTech Connect (OSTI)

    Gedeon, S.R.

    1997-04-11T23:59:59.000Z

    Spent light water reactor fuel is to be moved out of the 324 Building. It is anticipated that intact fuel assemblies will be loaded in a modified Multi-Canister Overpack Canister, which in turn will be placed in an Overpack Transportation Cask. An estimate of gamma ray dose rates from a transportation cask is desired.

  19. This A TM contains mate rial on the comparative analysis and de sign of the A LSEP Fuel Cask Thermal Shield. This report is submitted in accor-

    E-Print Network [OSTI]

    Rathbun, Julie A.

    of the fuel cask/LEM/SLA envelope was undertakei?- to determine the specific requirements for a thermal heat envelope was made to select the position an? orientation of the cask thermal shield. Figure 1 shows Thermal Shield. This report ·is submitted in accor- dance with MSC request generated during the System

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

    SciTech Connect (OSTI)

    ARD, K.E.

    2000-04-19T23:59:59.000Z

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

  1. 1 BASEMENT STORAGE 3 MICROSCOPE LAB

    E-Print Network [OSTI]

    Boonstra, Rudy

    MECHANICAL ROOM 13 SHOWER ROOMSAIR COMPRESSOR 14 NITROGEN STORAGE 15 DIESEL FUEL STORAGE 16 ACID NEUT. TANK NMR RELAXOMETER ROOM 13 LARGE MEETING ROOM (INCL. KITCHEN) 14 STIPEND / VISITOR OFFICE 15 GRAD OFFICE ROOM / TECH OFFICE 5 ELECTRICAL CLOSET 6 NMR RELAXOMETER ROOM 7 DRY SOLVENT ROOM 8 MEETING ROOM

  2. Drying studies for corroded DOE aluminum plate fuels

    SciTech Connect (OSTI)

    Lords, R.E.; Windes, W.E. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Crepeau, J.C.; Sidwell, R.W. [Idaho Univ., Idaho Falls, ID (United States) Dept. of Mechanical Engineering

    1996-05-01T23:59:59.000Z

    The Idaho National Engineering Laboratory (INEL) currently stores a wide variety of spent nuclear fuel. The fuel was originally intended to be stored underwater for a short period of thermal cooling, then removed and reprocessed. However, it has been stored underwater for much longer thank originally anticipated. During this time dust and airborne desert soil have entered the oldest INEL pool, accumulating on the fuel. Also, the aluminum fuel cladding has corroded compromising the exposed surfaces of the fuel. Plans are now underway to move some the the more vulnerable aluminum plate type fuels into dry storage in an existing vented and filtered fuel storage facility. In preparation for dry storage of the fuel a drying and canning station is being built at the INEL. The two primary objectives of this facility are to determine the influence of corrosion products on the drying process and to establish temperature distribution inside the canister during heating.

  3. Evaluation of postharvest quality of onion varieties during storage

    E-Print Network [OSTI]

    Rajapakse, Nihal Chandrakumara

    1983-01-01T23:59:59.000Z

    . However, this process should not cause excessive shrinkage due to removal of moisture from the interior of the bulb (47). Suitability of onions for storage is generally reached when the neck is tight and outer scales are dry (37, 38). This condition... of artificial curing and the need for dry outer scales before storage, have led to the study of preharvest foliar 12 desiccants on onions (19, 47, 49). Higher storage losses have been consistantly reported with the use of preharvest foliar desiccants...

  4. Self-protection in dry recycle technologies

    SciTech Connect (OSTI)

    Hannum, W.H.; Wade, D.; Stanford, G.

    1995-12-01T23:59:59.000Z

    In response to the INFCE conclusions, the U.S. undertook development of a new dry fuel cycle. Dry recycle processes have been demonstrated to be feasible. Safeguarding such fuel cycles will be dramatically simpler than the PUREX fuel cycle. At every step of the processes, the materials meet the {open_quotes}spent-fuel standard.{close_quotes} The scale is compatible with collocation of power reactors and their recycle facility, eliminating off-site transportation and storage of plutonium-bearing materials. Material diverted either covertly or overtly would be difficult (relative to material available by other means) to process into weapons feedstock.

  5. Scoping design analyses for optimized shipping casks containing 1-, 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect (OSTI)

    Bucholz, J.A.

    1983-01-01T23:59:59.000Z

    This report details many of the interrelated considerations involved in optimizing large Pb, Fe, or U-metal spent fuel shipping casks containing 1, 2, 3, 5, 7, or 10-year-old PWR fuel assemblies. Scoping analyses based on criticality, shielding, and heat transfer considerations indicate that some casks may be able to hold as many as 18 to 21 ten-year-old PWR fuel assemblies. In the criticality section, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. In another section, details of many n/..gamma.. shielding optimization studies are presented, including the optimal n/..gamma.. design points and the actual shielding requirements for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. Multigroup source terms based on ORIGEN2 calculations at these and other decay times are also included. Lastly, the numerical methods and experimental correlations used in the steady-state and transient heat transfer analyses are fully documented, as are pertinent aspects of the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. (While only casks for square, intact PWR fuel assemblies were considered in this study, the SCOPE code may also be used to design and analyze casks containing canistered spent fuel or other waste material. An abbreviated input data guide is included as an appendix).

  6. Testing of ethylene propylene seals for the GA-4/GA-9 casks

    SciTech Connect (OSTI)

    Boonstra, R.H.

    1993-08-01T23:59:59.000Z

    The primary O-ring seal of the GA-4 and GA-9 casks was tested for leakage with a full-scale mockup of the cask lid and flange. Tests were performed at temperatures of ambient, {minus}41{degrees}, 121{degrees}, and 193{degrees}C. Shim plates between the lid and flange simulated gaps caused by thermal distortion. The testing used a helium mass spectrometer leak detector (MSLD). Results showed that the primary seal was leaktight for all test conditions. Helium permeation through the seal began in 13--23 minutes for the ambient tests and in 1--2 minutes for the tests at elevated temperatures. After each test several hours of the pumping were typically required to reduce the MSLD background reading to an acceptable level for the next test, indicating that the seal had become saturated with helium. To verify that the test results showed permeation and not real leakage, several response checks were conducted in which a calibrated leak source was inserted in the detector line near the seal. When the leak source was activated the detector responded within seconds.

  7. Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

    SciTech Connect (OSTI)

    Bracey, William; Bondre, Jayant; Shelton, Catherine [Transnuclear, Inc., 7135 Minstrel Way Suite 300, Columbia MD 21045 (United States)] [Transnuclear, Inc., 7135 Minstrel Way Suite 300, Columbia MD 21045 (United States); Edmonds, Robert [AREVA Federal Services, 7207 IBM Drive, Charlotte NC 28262 (United States)] [AREVA Federal Services, 7207 IBM Drive, Charlotte NC 28262 (United States)

    2013-07-01T23:59:59.000Z

    The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In January 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to evaluate the solutions, and the alternative solutions. The complexity of the project is increasing with time (more fuel assemblies, new storage systems, deteriorating logistics infrastructure at some sites, etc.) but with the uncertainty on the final disposal path, flexibility and simplicity will be critical. (authors)

  8. Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

    SciTech Connect (OSTI)

    Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.; Mironov, B.S.; Kaplenkov, V.N.; Seredenko, A.V.; Saranchin, V.K.; Shulgin, A.S. [All-Russian Research Institute of Chemical Technology (ARRICT), Moscow (Russian Federation); Haire, M.J.; Forsberg, C.W. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2007-07-01T23:59:59.000Z

    This paper describes the results of a joint research program between the Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory in the United States to develop new radiation shielding materials for use in the construction of casks for spent nuclear fuel (SNF) and radioactive wastes. Research and development is underway to develop SNF storage, transport, and disposal casks using shielding made with two new depleted uranium dioxide (DUO{sub 2}) materials: a DUO{sub 2}-steel cermet, and, DUCRETE with DUAGG (DUO{sub 2} aggregate). Melting the DUO{sub 2} and allowing it to freeze will produce a near 100% theoretical density product and assures that the product produces no volatile materials upon subsequent heating. Induction cold-crucible melters (ICCM) are being developed for this specific application. An ICCM is, potentially, a high throughput low-cost process. Schematics of a pilot facility were developed for the production of molten DUO{sub 2} from DU{sub 3}O{sub 8} to produce granules <1 mm in diameter in a continuous mode of operation. Thermodynamic analysis was conducted for uranium-oxygen system in the temperature range from 300 to 4000 K in various gas mediums. Temperature limits of stability for various uranium oxides were determined. Experiments on melting DUO{sub 2} were carried out in a high frequency ICCM in a cold crucible with a 120 mm in diameter. The microstructure of molten DUO{sub 2} was studied and lattice parameters were determined. It was experimentally proved, and validated by X-ray analysis, that an opportunity exists to produce molten DUO{sub 2} from mixed oxides (primarily DU{sub 3}O{sub 8}) by reduction melting in ICCM. This will allow using DU{sub 3}O{sub 8} directly to make DUO{sub 2}-a separate unit operation to produce UO{sub 2} feed material is not needed. Experiments were conducted concerning the addition of alloying components, gadolinium et al. oxides, into the DUO{sub 2} melt while in the crucible. These additives improve neutron and gamma radiation shielding and operation properties of the final solids. Cermet samples of 50 wt % DUO{sub 2} were produced. (authors)

  9. Safety evaluation for packaging for the transport of K Basin sludge samples in the PAS-1 cask

    SciTech Connect (OSTI)

    SMITH, R.J.

    1998-11-17T23:59:59.000Z

    This safety evaluation for packaging authorizes the shipment of up to two 4-L sludge samples to and from the 325 Lab or 222-S Lab for characterization. The safety of this shipment is based on the current U.S. Department of Energy Certification of Compliance (CoC) for the PAS-1 cask, USA/9184/B(U) (DOE).

  10. NGLW RCRA Storage Study

    SciTech Connect (OSTI)

    R. J. Waters; R. Ochoa; K. D. Fritz; D. W. Craig

    2000-06-01T23:59:59.000Z

    The Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory contains radioactive liquid waste in underground storage tanks at the INTEC Tank Farm Facility (TFF). INTEC is currently treating the waste by evaporation to reduce the liquid volume for continued storage, and by calcination to reduce and convert the liquid to a dry waste form for long-term storage in calcine bins. Both treatment methods and activities in support of those treatment operations result in Newly Generated Liquid Waste (NGLW) being sent to TFF. The storage tanks in the TFF are underground, contained in concrete vaults with instrumentation, piping, transfer jets, and managed sumps in case of any liquid accumulation in the vault. The configuration of these tanks is such that Resource Conservation and Recovery Act (RCRA) regulations apply. The TFF tanks were assessed several years ago with respect to the RCRA regulations and they were found to be deficient. This study considers the configuration of the current tanks and the RCRA deficiencies identified for each. The study identifies four potential methods and proposes a means of correcting the deficiencies. The cost estimates included in the study account for construction cost; construction methods to minimize work exposure to chemical hazards, radioactive contamination, and ionizing radiation hazards; project logistics; and project schedule. The study also estimates the tank volumes benefit associated with each corrective action to support TFF liquid waste management planning.

  11. Energy Storage

    SciTech Connect (OSTI)

    Paranthaman, Parans

    2014-06-03T23:59:59.000Z

    ORNL Distinguished Scientist Parans Paranthaman is discovering new materials with potential for greatly increasing batteries' energy storage capacity and bring manufacturing back to the US.

  12. Energy Storage

    ScienceCinema (OSTI)

    Paranthaman, Parans

    2014-06-23T23:59:59.000Z

    ORNL Distinguished Scientist Parans Paranthaman is discovering new materials with potential for greatly increasing batteries' energy storage capacity and bring manufacturing back to the US.

  13. Terrestrial Water Storage

    E-Print Network [OSTI]

    Rodell, M; Chambers, D P; Famiglietti, Jay

    2013-01-01T23:59:59.000Z

    T. E. Reilly, 2002: Flow and storage in groundwater systems.storage ..2013: Global ocean storage of anthropogenic carbon.

  14. Stasis: Flexible Transactional Storage

    E-Print Network [OSTI]

    Sears, Russell C.

    2009-01-01T23:59:59.000Z

    storage . . . . . . . . . . . . . . . . . . . . . .example system based on log-structured storage 10.1 SystemA storage bottleneck. . . . . . . . . . . . . . . .

  15. Dry Process Electrode Fabrication

    Broader source: Energy.gov (indexed) [DOE]

    with good mechanical properties - Loading approaching targets - Process parameter optimization necessary to make thinner films with better density characteristics Images of dry...

  16. Transporting Dry Ice

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Requirements for Shipping Dry Ice IATA PI 904 Source: Reg of the Day from ERCweb 2006 Environmental Resource Center | 919-469-1585 | webmaster@ercweb.com http:...

  17. Sandia National Laboratories: DRI

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    DRI ECIS-Princeton Power Systems, Inc.: Demand Response Inverter On March 19, 2013, in DETL, Distribution Grid Integration, Energy, Energy Surety, Facilities, Grid Integration,...

  18. Cooking with Dry Spaghetti

    E-Print Network [OSTI]

    Anding, Jenna

    2008-12-09T23:59:59.000Z

    This fact sheet describes the nutritional value and safe storage of spaghetti, a commodity food. It also offers food preparation ideas....

  19. Silo Storage Preconceptual Design

    SciTech Connect (OSTI)

    Stephanie L. Austad; Patrick W. Bragassa; Kevin M Croft; David S Ferguson; Scott C Gladson; Annette L Shafer; John H Weathersby

    2012-09-01T23:59:59.000Z

    The National Nuclear Security Administration (NNSA) has a need to develop and field a low-cost option for the long-term storage of a variety of radiological material. The storage option’s primary requirement is to provide both environmental and physical protection of the materials. Design criteria for this effort require a low initial cost and minimum maintenance over a 50-year design life. In 1999, Argonne National Laboratory-West was tasked with developing a dry silo storage option for the BN-350 Spent Fuel in Aktau Kazakhstan. Argon’s design consisted of a carbon steel cylinder approximately 16 ft long, 18 in. outside diameter and 0.375 in. wall thickness. The carbon steel silo was protected from corrosion by a duplex coating system consisting of zinc and epoxy. Although the study indicated that the duplex coating design would provide a design life well in excess of the required 50 years, the review board was concerned because of the novelty of the design and the lack of historical use. In 2012, NNSA tasked Idaho National Laboratory (INL) with reinvestigating the silo storage concept and development of alternative corrosion protection strategies. The 2012 study, “Silo Storage Concepts, Cathodic Protection Options Study” (INL/EST-12-26627), concludes that the option which best fits the design criterion is a passive cathotic protection scheme, consisting of a carbon steel tube coated with zinc or a zinc-aluminum alloy encapsulated in either concrete or a cement grout. The hot dipped zinc coating option was considered most efficient, but the flame-sprayed option could be used if a thicker zinc coating was determined to be necessary.

  20. Hydrogen Storage

    Fuel Cell Technologies Publication and Product Library (EERE)

    This 2-page fact sheet provides a brief introduction to hydrogen storage technologies. Intended for a non-technical audience, it explains the different ways in which hydrogen can be stored, as well a

  1. Safety Issues Chemical Storage

    E-Print Network [OSTI]

    Cohen, Robert E.

    Safety Issues · Chemical Storage ·Store in compatible containers that are in good condition to store separately. #12;Safety Issues · Flammable liquid storage -Store bulk quantities in flammable storage cabinets -UL approved Flammable Storage Refrigerators are required for cold storage · Provide

  2. Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

    SciTech Connect (OSTI)

    Gotovchikov, Vitaly; Seredenko, V.A.; Shatalov, V.V.; Mironov, B.S.; Kaplenkov, V.N.; Seredenko, A.V.; Saranchin, V.K.; Shulgin, A.S.; Kalmakov, Danila [All-Russian Research Institute of Chemical Technology (ARRICT), Kashirskoe Shosse 33, Moscow 115230 (Russian Federation); Haire, M.J.; Forsberg, C.W. [Oak Ridge National Laboratory - ORNL, 1 Bethel Valley Rd, Oak Ridge, TN 37830 (United States)

    2007-07-01T23:59:59.000Z

    Available in abstract form only. Full text of publication follows: This paper describes the results of joint research program of Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory to develop new materials for build spent nuclear fuel (SNF) storage, transport, and disposal casks using shielding made with depleted uranium dioxide (DUO{sub 2}) in a DUO{sub 2}-steel cermet or a DUCRETE with DUAGG (DUO{sub 2} aggregate) with selective additives in cement matrix. The preparation of DUO{sub 2} particles and aggregates for shielding could be produced from technologies that are extrapolated from the costly multi-step nuclear fuel pellet technologies. Melting the DUO{sub 2} and allowing it to freeze will produce a product near 100% theoretical density and assure that the product produces no volatile materials upon subsequent heating. Melting is a one step process that provides an opportunity to include additives in the DUO{sub 2} to modify its chemical or nuclear properties. The proposed work is directed to develop cold-wall induction heated melters (ICCM) for this specific application. Experiments on melting DUO{sub 2} were carried out in high frequency ICCM with cold crucible. It was experimentally proved an opportunity to produce molten DUO{sub 2} from mixed oxides (DU{sub 3}O{sub 8}) by reducing melting in ICCM. This will allow using DU{sub 3}O{sub 8} generated in direct conversion of depleted uranium hexafluoride as source material for melted and granulated DUO{sub 2} production. Experiments on the addition of alloying components - gadolinium oxide and others into DUO{sub 2} melt while in crucible to improve neutron and gamma radiation-shielding and operation properties of the final solids were carried out. (authors)

  3. Dry Process Electrode Fabrication

    Broader source: Energy.gov (indexed) [DOE]

    Ratecapacity match cathode 12 8. Down-select low cost anode process 50% vs baseline capex + opex 13 9. Scale cathode film to support task 16 10 m 17 10. Lab prototype cell dry...

  4. Freeze drying method

    DOE Patents [OSTI]

    Coppa, Nicholas V. (Malvern, PA); Stewart, Paul (Youngstown, NY); Renzi, Ernesto (Youngstown, NY)

    1999-01-01T23:59:59.000Z

    The present invention provides methods and apparatus for freeze drying in which a solution, which can be a radioactive salt dissolved within an acid, is frozen into a solid on vertical plates provided within a freeze drying chamber. The solid is sublimated into vapor and condensed in a cold condenser positioned above the freeze drying chamber and connected thereto by a conduit. The vertical positioning of the cold condenser relative to the freeze dryer helps to help prevent substances such as radioactive materials separated from the solution from contaminating the cold condenser. Additionally, the system can be charged with an inert gas to produce a down rush of gas into the freeze drying chamber to also help prevent such substances from contaminating the cold condenser.

  5. Freeze drying apparatus

    DOE Patents [OSTI]

    Coppa, Nicholas V. (Malvern, PA); Stewart, Paul (Youngstown, NY); Renzi, Ernesto (Youngstown, NY)

    2001-01-01T23:59:59.000Z

    The present invention provides methods and apparatus for freeze drying in which a solution, which can be a radioactive salt dissolved within an acid, is frozen into a solid on vertical plates provided within a freeze drying chamber. The solid is sublimated into vapor and condensed in a cold condenser positioned above the freeze drying chamber and connected thereto by a conduit. The vertical positioning of the cold condenser relative to the freeze dryer helps to help prevent substances such as radioactive materials separated from the solution from contaminating the cold condenser. Additionally, the system can be charged with an inert gas to produce a down rush of gas into the freeze drying chamber to also help prevent such substances from contaminating the cold condenser.

  6. Assessment of an active dry barrier for a landfill cover system

    SciTech Connect (OSTI)

    Stormont, J.C. [Sandia National Labs., Albuquerque, NM (United States); Ankeny, M.D.; Burkhard, M.E.; Tansey, M.K.; Kelsey, J.A. [Stephens (Daniel B.) and Associates, Inc., Albuquerque, NM (United States)

    1994-03-01T23:59:59.000Z

    A dry barrier is a layer of geologic material that is dried by air flow. An active dry barrier system can be designed, installed, and operated as part of a landfill cover system. An active system uses blowers and fans to move air through a high-permeability layer within the cover system. Depending principally on the air-flow rate, it is possible for a dry barrier to remove enough water to substantially reduce the likelihood of water percolating through the cover system. If a material with a relatively great storage capacity, such as processed tuff, is used as the coarse layer, then the efficiency of the dry barrier will be increased.

  7. 1 Copyright 2012 by ASME Proceedings of the ASME 2012 Pressure Vessels & Piping Division Conference

    E-Print Network [OSTI]

    Giurgiutiu, Victor

    requirements for monitoring and inspection of dry storage systems as part of aging management plans. FIGURE 1.S. Department of Energy (DOE) as a high priority cross-cutting need. Monitoring is necessary to determine: DRY CASK STORAGE SYSTEM FOR SPENT NUCLEAR FUEL Structural health monitoring offers the solution

  8. Targets for on-board hydrogen storage systems: Current R&D focus is on 2010 Targets

    E-Print Network [OSTI]

    storage system) 98% (dry basis) Useful constants: 0.2778kWh/MJ, ~33.3kWh/gal gasoline equivalent. #12Targets for on-board hydrogen storage systems: Current R&D focus is on 2010 Targets Table 1. DOE Technical Targets: On-Board Hydrogen Storage Systemsa, b, c Storage Parameter Units 2007* 2010 2015 Usable

  9. NUHOWS - Storage and Transportation of Irradiated Reactor Components in Large Packages - 13439

    SciTech Connect (OSTI)

    Rae, Glen A. [Transnuclear, Inc., 7135 Minstrel Way, Columbia, MD 21045 (United States)] [Transnuclear, Inc., 7135 Minstrel Way, Columbia, MD 21045 (United States)

    2013-07-01T23:59:59.000Z

    Most irradiated reactor components (hardware such as Control Rod Blades, Fuel Channels, Poison Curtains, etc.) generated at reactors previously required significant processing for size reduction due to the available transportation casks not being physically capable of containing unprocessed material. As of July 1, 2008, disposal for this typical waste class (B and C) became inaccessible (for the major part of the nation) due to the Barnwell, SC disposal facility being closed to all but its three compact states (CT, NJ and SC). Currently in the United States, most facilities are storing their irradiated hardware on-site in the spent fuel pools. Until recently with the opening of the Waste Control Specialists' Texas disposal facility, utilities faced the challenges of spent fuel pool space and capacity management. However, even with WCS's disposal availability, the site currently has annual Curie limitations for disposal, which will continue to promote interim on-site storage until such time as disposal is available. In response, Transnuclear Inc., (TN) an AREVA company, proceeded with designing a new large Radioactive Waste Container (RWC) that can be used to package irradiated hardware without the need for significant processing. The design features of the RWC allows for intermittent loadings of the hardware for better packaging efficiency, higher packaging density, space savings and reduced cost. This RWC is also compatible with TN's on-site modular vault storage system. Once completely loaded, the RWC can be transported to an on-site storage facility, an off-site storage facility and/or an available disposal facility. To accommodate the transportation, TN has designed a large transportation cask, the MP197HB. As the original design was for transporting fuel, it contains the necessary shielding to allow for the transport of unprocessed irradiated reactor components, while significantly reducing the amount of irradiated hardware shipments required with the use of the new RWC. This paper provides information on the unique design features of the RWC, storage module vaults, MP197HB Transportation Cask and cost saving benefits of using the large RWC for packaging, storage, transport and disposal. (authors)

  10. AQUIFER THERMAL ENERGY STORAGE

    E-Print Network [OSTI]

    Tsang, C.-F.

    2011-01-01T23:59:59.000Z

    aquifers for thermal energy storage. Problems outlined aboveModeling of Thermal Energy Storage in Aquifers," Proceed-ings of Aquifer Thermal Energy Storage Workshop, Lawrence

  11. SUPERCONDUCTING MAGNETIC ENERGY STORAGE

    E-Print Network [OSTI]

    Hassenzahl, W.

    2011-01-01T23:59:59.000Z

    Superconducting 30-MJ Energy Storage Coil", Proc. 19 80 ASC,Superconducting Magnetic Energy Storage Plant", IEEE Trans.SlIperconducting Magnetic Energy Storage Unit", in Advances

  12. AQUIFER THERMAL ENERGY STORAGE

    E-Print Network [OSTI]

    Tsang, C.-F.

    2011-01-01T23:59:59.000Z

    aquifers for thermal energy storage. Problems outlined abovean Aquifer Used for Hot Water Storage: Digital Simulation ofof Aquifer Systems for Cyclic Storage of Water," of the Fall

  13. AQUIFER THERMAL ENERGY STORAGE

    E-Print Network [OSTI]

    Tsang, C.-F.

    2011-01-01T23:59:59.000Z

    using aquifers for thermal energy storage. Problems outlinedmatical Modeling of Thermal Energy Storage in Aquifers,"ings of Aquifer Thermal Energy Storage Workshop, Lawrence

  14. AQUIFER THERMAL ENERGY STORAGE

    E-Print Network [OSTI]

    Tsang, C.-F.

    2011-01-01T23:59:59.000Z

    using aquifers for thermal energy storage. Problems outlinedmatical Modeling of Thermal Energy Storage in Aquifers,"Proceed- ings of Aquifer Thermal Energy Storage Workshop,

  15. Stasis: Flexible Transactional Storage

    E-Print Network [OSTI]

    Sears, Russell C.

    2009-01-01T23:59:59.000Z

    Stasis: Flexible Transactional Storage by Russell C. Sears AR. Larson Fall 2009 Stasis: Flexible Transactional StorageC. Sears Abstract Stasis: Flexible Transactional Storage by

  16. Spent fuel storage requirements 1993--2040

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges.

  17. Cooking with Dried Potatoes

    E-Print Network [OSTI]

    Anding, Jenna

    2008-12-09T23:59:59.000Z

    make a tasty vegetable dish. For added flavor, you can add salt and pepper along with small amounts of grated cheese, margarine or butter. Be careful: Adding large amounts of cheese, butter or margarine can turn a low-fat vegetable, such as potatoes..., into a high-fat dish. How to store them Store packages of dried potatoes in a cool, dry, place. After the package is opened, store the potatoes in an airtight container. Store cooked potatoes in a covered dish in the refrigerator. Use within 3 days...

  18. Storage Rings

    SciTech Connect (OSTI)

    Fischer, W.

    2011-01-01T23:59:59.000Z

    Storage rings are circular machines that store particle beams at a constant energy. Beams are stored in rings without acceleration for a number of reasons (Tab. 1). Storage rings are used in high-energy, nuclear, atomic, and molecular physics, as well as for experiments in chemistry, material and life sciences. Parameters for storage rings such as particle species, energy, beam intensity, beam size, and store time vary widely depending on the application. The beam must be injected into a storage ring but may not be extracted (Fig. 1). Accelerator rings such as synchrotrons are used as storage rings before and after acceleration. Particles stored in rings include electrons and positrons; muons; protons and anti-protons; neutrons; light and heavy, positive and negative, atomic ions of various charge states; molecular and cluster ions, and neutral polar molecules. Spin polarized beams of electrons, positrons, and protons were stored. The kinetic energy of the stored particles ranges from 10{sup -6} eV to 3.5 x 10{sup 12} eV (LHC, 7 x 10{sup 12} eV planned), the number of stored particles from one (ESR) to 1015 (ISR). To store beam in rings requires bending (dipoles) and transverse focusing (quadrupoles). Higher order multipoles are used to correct chromatic aberrations, to suppress instabilities, and to compensate for nonlinear field errors of dipoles and quadrupoles. Magnetic multipole functions can be combined in magnets. Beams are stored bunched with radio frequency systems, and unbunched. The magnetic lattice and radio frequency system are designed to ensure the stability of transverse and longitudinal motion. New technologies allow for better storage rings. With strong focusing the beam pipe dimensions became much smaller than previously possible. For a given circumference superconducting magnets make higher energies possible, and superconducting radio frequency systems allow for efficient replenishment of synchrotron radiation losses of large current electron or positron beams. Storage rings have instrumentation to monitor the electrical and mechanical systems, and the beam quality. Computers are used to control the operation. Large storage rings have millions of control points from all systems. The time dependent beam intensity I(t) can often be approximated by an exponential function I(t) = I(0) exp(-t/{tau}) (1) where the decay time {tau} and, correspondingly, the store time ranges from a few turns to 10 days (ISR). {tau} can be dominated by a variety of effects including lattice nonlinearities, beam-beam, space charge, intrabeam and Touschek scattering, interaction with the residual gas or target, or the lifetime of the stored particle. In this case, the beam lifetime measurement itself can be the purpose of a storage ring experiment. The main consideration in the design of a storage ring is the preservation of the beam quality over the store length. The beam size and momentum spread can be reduced through cooling, often leading to an increase in the store time. For long store times vacuum considerations are important since the interaction rate of the stored particles with the residual gas molecules is proportional to the pressure, and an ultra-high vacuum system may be needed. Distributed pumping with warm activated NEG surfaces or cold surfaces in machines with superconducting magnets are ways to provide large pumping speeds and achieve low pressures even under conditions with dynamic gas loads. The largest application of storage rings today are synchrotron light sources, of which about 50 exist world wide. In experiments where the beam collides with an internal target or another beam, a storage ring allows to re-use the accelerated beam many times if the interaction with the target is sufficiently small. In hadron collider and ion storage rings store times of many hours or even days are realized, corresponding to up to 1011 turns and thereby target passages. Ref. [3] is the first proposal for a collider storage ring. A number of storage rings exist where the beam itself or its decay products are the object of s

  19. Cool Storage Performance

    E-Print Network [OSTI]

    Eppelheimer, D. M.

    1985-01-01T23:59:59.000Z

    . This article covers three thermal storage topics. The first section catalogs various thermal storage systems and applications. Included are: load shifting and load leveling, chilled water storage systems, and ice storage systems using Refrigerant 22 or ethylene...

  20. Monitoring groundwater storage changes in the highly1 seasonal humid tropics: validation of GRACE measurements2

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 Monitoring groundwater storage changes in the highly1 seasonal humid tropics: validation of GRACE the seasonality and trend in groundwater storage associated with intensive groundwater19 abstraction for dry to 2007) groundwater storage changes21 (GWS) correlate well (r=0.77 to 0.93, p-value

  1. Cooling Dry Cows

    E-Print Network [OSTI]

    Stokes, Sandra R.

    2000-07-17T23:59:59.000Z

    , little work has been done on the responses of cooling cows in this period. The dry period is particularly crucial because it involves regen- eration of the mammary gland and rapid fetal growth. This is also when follicles begin develop- ing and maturing...

  2. AQUIFER THERMAL ENERGY STORAGE

    E-Print Network [OSTI]

    Tsang, C.-F.

    2011-01-01T23:59:59.000Z

    of Discharge Using Ground- Water Storage," Transactions1971. "Storage of Solar Energy in a Sandy-Gravel Ground,"

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

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Suter, Jonathan D.; Jones, Anthony M.

    2014-09-12T23:59:59.000Z

    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.

  4. Research on Farm Drying and Storage of Sorghum Grain.

    E-Print Network [OSTI]

    Redlinger, L. M.; Davenport, M. G.; Sorenson Jr., J. W.; Kline, G. L.; Aldred, W. H.

    1957-01-01T23:59:59.000Z

    Deterioration. Cereal Chemistry. Vol. 34, No. 4, pp. 226-33. July 1957. (3) Neal, E. M., R. A. Hall and J. H. Jones. Live and Dead Germ Sorghum Grain in Steer Fattening Rations. Progress Report 1702. Texas Agricultural Experiment Station. July 1954. (4... of the following individuals: M. M. Garcia, of Substation No. 4, Beaumont, Texas, for his assistance in conducting the tests : M. D. Whitehead, formerly associate professor, Department of Plant Physiology and Path- ology, College Station, Texas, for conducting...

  5. Occupational dose estimates for a monitored retrievable storage facility

    SciTech Connect (OSTI)

    Harty, R.; Stoetzel, G.A.

    1986-06-01T23:59:59.000Z

    Occupational doses were estimated for radiation workers at the monitored retrievable storage (MRS) facility. This study provides an estimate of the occupational dose based on the current MRS facility design, examines the extent that various design parameters and assumptions affect the dose estimates, and identifies the areas and activities where exposures can be reduced most effectively. Occupational doses were estimated for both the primary storage concept and the alternate storage concept. The dose estimates indicate the annual dose to all radiation workers will be below the 5 rem/yr federal dose equivalent limit. However, the estimated dose to most of the receiving and storage crew (the workers responsible for the receipt, storage, and surveillance of the spent fuel and its subsequent retrieval), to the crane maintenance technicians, and to the cold and remote maintenance technicians is above the design objective of 1 rem/yr. The highest annual dose is received by the riggers (4.7 rem) in the receiving and storage crew. An indication of the extent to which various design parameters and assumptions affect the dose estimates was obtained by changing various design-based assumptions such as work procedures, background dose rates in radiation zones, and the amount of fuel received and stored annually. The study indicated that a combination of remote operations, increased shielding, and additional personnel (for specific jobs) or changes in operating procedures will be necessary to reduce worker doses below 1.0 rem/yr. Operations that could be made at least partially remote include the removal and replacement of the tiedowns, impact limiters, and personnel barriers from the shipping casks and the removal or installation of the inner closure bolts. Reductions of the background dose rates in the receiving/shipping and the transfer/discharge areas may be accomplished with additional shielding.

  6. Storage System and IBM System Storage

    E-Print Network [OSTI]

    IBM® XIV® Storage System and IBM System Storage® SAN Volume Controller deliver high performance and smart management for SAP® landscapes IBM SAP International Competence Center #12;"The combination of the XIV Storage System and SAN Volume Controller gives us a smarter way to manage our storage. If we need

  7. Drying of fiber webs

    DOE Patents [OSTI]

    Warren, David W. (9253 Glenoaks Blvd., Sun Valley, CA 91352)

    1997-01-01T23:59:59.000Z

    A process and an apparatus for high-intensity drying of fiber webs or sheets, such as newsprint, printing and writing papers, packaging paper, and paperboard or linerboard, as they are formed on a paper machine. The invention uses direct contact between the wet fiber web or sheet and various molten heat transfer fluids, such as liquified eutectic metal alloys, to impart heat at high rates over prolonged durations, in order to achieve ambient boiling of moisture contained within the web. The molten fluid contact process causes steam vapor to emanate from the web surface, without dilution by ambient air; and it is differentiated from the evaporative drying techniques of the prior industrial art, which depend on the uses of steam-heated cylinders to supply heat to the paper web surface, and ambient air to carry away moisture, which is evaporated from the web surface. Contact between the wet fiber web and the molten fluid can be accomplished either by submersing the web within a molten bath or by coating the surface of the web with the molten media. Because of the high interfacial surface tension between the molten media and the cellulose fiber comprising the paper web, the molten media does not appreciately stick to the paper after it is dried. Steam generated from the paper web is collected and condensed without dilution by ambient air to allow heat recovery at significantly higher temperature levels than attainable in evaporative dryers.

  8. Drying of fiber webs

    DOE Patents [OSTI]

    Warren, D.W.

    1997-04-15T23:59:59.000Z

    A process and an apparatus are disclosed for high-intensity drying of fiber webs or sheets, such as newsprint, printing and writing papers, packaging paper, and paperboard or linerboard, as they are formed on a paper machine. The invention uses direct contact between the wet fiber web or sheet and various molten heat transfer fluids, such as liquefied eutectic metal alloys, to impart heat at high rates over prolonged durations, in order to achieve ambient boiling of moisture contained within the web. The molten fluid contact process causes steam vapor to emanate from the web surface, without dilution by ambient air; and it is differentiated from the evaporative drying techniques of the prior industrial art, which depend on the uses of steam-heated cylinders to supply heat to the paper web surface, and ambient air to carry away moisture, which is evaporated from the web surface. Contact between the wet fiber web and the molten fluid can be accomplished either by submersing the web within a molten bath or by coating the surface of the web with the molten media. Because of the high interfacial surface tension between the molten media and the cellulose fiber comprising the paper web, the molten media does not appreciatively stick to the paper after it is dried. Steam generated from the paper web is collected and condensed without dilution by ambient air to allow heat recovery at significantly higher temperature levels than attainable in evaporative dryers. 6 figs.

  9. Gas storage materials, including hydrogen storage materials

    DOE Patents [OSTI]

    Mohtadi, Rana F; Wicks, George G; Heung, Leung K; Nakamura, Kenji

    2014-11-25T23:59:59.000Z

    A material for the storage and release of gases comprises a plurality of hollow elements, each hollow element comprising a porous wall enclosing an interior cavity, the interior cavity including structures of a solid-state storage material. In particular examples, the storage material is a hydrogen storage material, such as a solid state hydride. An improved method for forming such materials includes the solution diffusion of a storage material solution through a porous wall of a hollow element into an interior cavity.

  10. Gas storage materials, including hydrogen storage materials

    DOE Patents [OSTI]

    Mohtadi, Rana F; Wicks, George G; Heung, Leung K; Nakamura, Kenji

    2013-02-19T23:59:59.000Z

    A material for the storage and release of gases comprises a plurality of hollow elements, each hollow element comprising a porous wall enclosing an interior cavity, the interior cavity including structures of a solid-state storage material. In particular examples, the storage material is a hydrogen storage material such as a solid state hydride. An improved method for forming such materials includes the solution diffusion of a storage material solution through a porous wall of a hollow element into an interior cavity.

  11. Microwavable thermal energy storage material

    DOE Patents [OSTI]

    Salyer, Ival O. (Dayton, OH)

    1998-09-08T23:59:59.000Z

    A microwavable thermal energy storage material is provided which includes a mixture of a phase change material and silica, and a carbon black additive in the form of a conformable dry powder of phase change material/silica/carbon black, or solid pellets, films, fibers, moldings or strands of phase change material/high density polyethylene/ethylene-vinyl acetate/silica/carbon black which allows the phase change material to be rapidly heated in a microwave oven. The carbon black additive, which is preferably an electrically conductive carbon black, may be added in low concentrations of from 0.5 to 15% by weight, and may be used to tailor the heating times of the phase change material as desired. The microwavable thermal energy storage material can be used in food serving applications such as tableware items or pizza warmers, and in medical wraps and garments.

  12. Microwavable thermal energy storage material

    DOE Patents [OSTI]

    Salyer, I.O.

    1998-09-08T23:59:59.000Z

    A microwavable thermal energy storage material is provided which includes a mixture of a phase change material and silica, and a carbon black additive in the form of a conformable dry powder of phase change material/silica/carbon black, or solid pellets, films, fibers, moldings or strands of phase change material/high density polyethylene/ethylene vinyl acetate/silica/carbon black which allows the phase change material to be rapidly heated in a microwave oven. The carbon black additive, which is preferably an electrically conductive carbon black, may be added in low concentrations of from 0.5 to 15% by weight, and may be used to tailor the heating times of the phase change material as desired. The microwavable thermal energy storage material can be used in food serving applications such as tableware items or pizza warmers, and in medical wraps and garments. 3 figs.

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

    SciTech Connect (OSTI)

    KRAHN, D.E.

    2000-08-08T23:59:59.000Z

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

  14. 2010 Dry Bean Research Report

    E-Print Network [OSTI]

    2010 Dry Bean Research Report Assessment of Narrow Row Technology Michigan Dry Edible Bean Production RESEARCH ADVISORY BOARD #12;The Michigan Bean Commission was awarded a grant from the MDA Technology for the Michigan Dry Bean Industry". Expected outcomes from this project are: 1. Identification

  15. 2012 Dry Bean Research Report

    E-Print Network [OSTI]

    2012 Dry Bean Research Report Assessment of Narrow Row Technology Michigan Dry Edible Bean Production Research Advisory Board #12;The Michigan Bean Commission was awarded a grant from the MDA Technology for the Michigan Dry Bean Industry". Expected outcomes from this project are: 1. Identification

  16. Analysis of radiation doses from operation of postulated commercial spent fuel transportation systems: Analysis of a system containing a monitored retrievable storage facility. Addendum 1

    SciTech Connect (OSTI)

    Smith, R.I.; Daling, P.M. [Pacific Northwest Lab., Richland, WA (United States); Faletti, D.W. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-04-01T23:59:59.000Z

    This addendum report extends the original study of the estimated radiation doses to the public and to workers resulting from transporting spent nuclear fuel from commercial nuclear power reactor stations through the federal waste management system (FWMS), to a system that contains a monitored retrievable storage (MRS) facility. The system concepts and designs utilized herein are consistent with those used in the original study (circa 1985--1987). Because the FWMS design is still evolving, the results of these analyses may no longer apply to the design for casks and cask handling systems that are currently being considered. Four system scenarios are examined and compared with the reference No-MRS scenario (all spent fuel transported directly from the reactors to the western repository in standard-capacity truck and rail casks). In Scenarios 1 and 2, an MRS facility is located in eastern United States and ships either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters. In Scenarios 3 and 4, an MRS facility is located in the western United States and ship either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters.

  17. Solid-State Hydrogen Storage: Storage Capacity,Thermodynamics...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Hydrogen Storage: Storage Capacity,Thermodynamics and Kinetics. Solid-State Hydrogen Storage: Storage Capacity,Thermodynamics and Kinetics. Abstract: Solid-state reversible...

  18. Sandia National Laboratories: Energy Storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Storage Sandian Spoke at the New York Energy Storage Expo On December 12, 2014, in Energy, Energy Storage, Energy Storage Systems, Grid Integration, Infrastructure Security, News,...

  19. Sandia National Laboratories: hydrogen storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    storage Energy Department Awards 7M to Advance Hydrogen Storage Systems On June 12, 2014, in CRF, Energy, Energy Storage, Energy Storage Systems, Facilities, Infrastructure...

  20. Method of drying articles

    DOE Patents [OSTI]

    Janney, M.A.; Kiggans, J.O. Jr.

    1999-03-23T23:59:59.000Z

    A method of drying a green particulate article includes the steps of: (a) Providing a green article which includes a particulate material and a pore phase material, the pore phase material including a solvent; and (b) contacting the green article with a liquid desiccant for a period of time sufficient to remove at least a portion of the solvent from the green article, the pore phase material acting as a semipermeable barrier to allow the solvent to be sorbed into the liquid desiccant, the pore phase material substantially preventing the liquid desiccant from entering the pores. 3 figs.

  1. Method of drying articles

    DOE Patents [OSTI]

    Janney, Mark A. (Knoxville, TN); Kiggans, Jr., James O. (Oak Ridge, TN)

    1999-01-01T23:59:59.000Z

    A method of drying a green particulate article includes the steps of: a. Providing a green article which includes a particulate material and a pore phase material, the pore phase material including a solvent; and b. contacting the green article with a liquid desiccant for a period of time sufficient to remove at least a portion of the solvent from the green article, the pore phase material acting as a semipermeable barrier to allow the solvent to be sorbed into the liquid desiccant, the pore phase material substantially preventing the liquid desiccant from entering the pores.

  2. Photon Storage Cavities

    E-Print Network [OSTI]

    Kim, K.-J.

    2008-01-01T23:59:59.000Z

    Sessler, "Analysis of Photon Storage Cavities for a Free-configuration of coupled storage cavity and PEL cavity. TheFig. 2. A ring resonator storage cavity coupled through a

  3. Seasonal thermal energy storage

    SciTech Connect (OSTI)

    Allen, R.D.; Kannberg, L.D.; Raymond, J.R.

    1984-05-01T23:59:59.000Z

    This report describes the following: (1) the US Department of Energy Seasonal Thermal Energy Storage Program, (2) aquifer thermal energy storage technology, (3) alternative STES technology, (4) foreign studies in seasonal thermal energy storage, and (5) economic assessment.

  4. Integrated Ingredients Dehydrated Agricultural Drying Low Temperature...

    Open Energy Info (EERE)

    Ingredients Dehydrated Agricultural Drying Low Temperature Geothermal Facility Jump to: navigation, search Name Integrated Ingredients Dehydrated Agricultural Drying Low...

  5. RH-TRU Waste Shipments from Battelle Columbus Laboratories to the Hanford Nuclear Facility for Interim Storage

    SciTech Connect (OSTI)

    Eide, J.; Baillieul, T. A.; Biedscheid, J.; Forrester, T,; McMillan, B.; Shrader, T.; Richterich, L.

    2003-02-26T23:59:59.000Z

    Battelle Columbus Laboratories (BCL), located in Columbus, Ohio, must complete decontamination and decommissioning (D&D) activities for nuclear research buildings and grounds by 2006, as directed by Congress. Most of the resulting waste (approximately 27 cubic meters [m3]) is remote-handled (RH) transuranic (TRU) waste destined for disposal at the Waste Isolation Pilot Plant (WIPP). The BCL, under a contract to the U.S. Department of Energy (DOE) Ohio Field Office, has initiated a plan to ship the TRU waste to the DOE Hanford Nuclear Facility (Hanford) for interim storage pending the authorization of WIPP for the permanent disposal of RH-TRU waste. The first of the BCL RH-TRU waste shipments was successfully completed on December 18, 2002. This BCL shipment of one fully loaded 10-160B Cask was the first shipment of RH-TRU waste in several years. Its successful completion required a complex effort entailing coordination between different contractors and federal agencies to establish necessary supporting agreements. This paper discusses the agreements and funding mechanisms used in support of the BCL shipments of TRU waste to Hanford for interim storage. In addition, this paper presents a summary of the efforts completed to demonstrate the effectiveness of the 10-160B Cask system. Lessons learned during this process are discussed and may be applicable to other TRU waste site shipment plans.

  6. Storage and IO Technology

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Burst Buffer User Defined Images Archive Home R & D Storage and IO Technologies Storage and IO Technologies Burst Buffer NVRAM and Burst Buffer Use Cases In collaboration...

  7. NERSC HPSS Storage Statistics

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Storage Trends and Summaries Storage by Scientific Discipline Troubleshooting IO Resources for Scientific Applications at NERSC Optimizing IO performance on the Lustre file...

  8. SUPERCONDUCTING MAGNETIC ENERGY STORAGE

    E-Print Network [OSTI]

    Hassenzahl, W.

    2011-01-01T23:59:59.000Z

    to MW/40 MWI-IR Battery Energy Storage Facility", proc. 23rdcompressed air, and battery energy storage are all only 65

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

    SciTech Connect (OSTI)

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

    1984-04-01T23:59:59.000Z

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

  10. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    SciTech Connect (OSTI)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01T23:59:59.000Z

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO/sub 2/ oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO/sub 2/ pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs.

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

    SciTech Connect (OSTI)

    Dana, W.P.

    1995-12-01T23:59:59.000Z

    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

  12. LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE

    SciTech Connect (OSTI)

    Dunn, K. [Savannah River National Laboratory; Bellamy, S. [Savannah River National Laboratory; Daugherty, W. [Savannah River National Laboratory; Sindelar, R. [Savannah River National Laboratory; Skidmore, E. [Savannah River National Laboratory

    2013-08-18T23:59:59.000Z

    Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintain integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.

  13. 2013 Dry Bean Research Report

    E-Print Network [OSTI]

    Page 1 2013 Dry Bean Research Report Black Bean Color Retention and White Mold Control in Narrow Row Production Systems Michigan Dry Edible Bean Production Research Advisory Board #12;Page 2 The Michigan Bean Commission was awarded a grant from the MDARD Specialty Crop Block Grant Program-Farm Bill

  14. Measurements of Fundamental Fluid Physics of SNF Storage Canisters

    SciTech Connect (OSTI)

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

    2001-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2002-01-29T23:59:59.000Z

    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.

  16. Making the Case For Safe Storage of Used Nuclear Fuel For Extended Periods of Time: Combining Near-Term Experiments and Analyses with Longer-Term Confirmatory Demonstrations

    SciTech Connect (OSTI)

    Sorenson, Ken B.; Hanson, Brady D.

    2013-08-25T23:59:59.000Z

    The need for extended storage of used nuclear fuel is increasing globally as disposition schedules for used fuel are pushed further into the future. This is creating a situation where dry storage of used fuel may need to be extended beyond normal regulatory licensing periods. While it is generally accepted that used fuel in dry storage will remain in a safe condition, there is little data that demonstrate used fuel performance in dry storage environments for long periods of time. This is especially true for high burnup used fuel.

  17. High-intensity drying processes: Impulse drying. Annual report

    SciTech Connect (OSTI)

    Orloff, D.I.; Phelan, P.M.

    1993-12-01T23:59:59.000Z

    Experiments were conducted on a sheet-fed pilot-scale shoe press to compare impulse drying and double-felted pressing. Both an IPST (Institute of Paper Science and Technology) ceramic coated and Beloit Type A press roll were evaluated for lienrboard sheet structures having a wide range of z-direction permeability. Purpose was to find ways of correcting sheet sticking problems observed in previous pilot-scale shoe press experiments. Results showed that impulse drying was superior to double felted pressing in both press dryness and in important paper physical properties. Impulse drying critical temperature was found to depend on specific surface of the heated layer of the sheet, thermal properties of the press roll surface, and choice of felt. Impulse drying of recycled and two-ply liner was demonstrated for both Southern Pile and Douglas fir-containing furnishes.

  18. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations; Yucca Mountain Site Characterization Project

    SciTech Connect (OSTI)

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States)

    1991-12-01T23:59:59.000Z

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 {times} 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990.

  19. CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

    SciTech Connect (OSTI)

    Mickalonis, J.

    2014-06-01T23:59:59.000Z

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  20. Structure Optimization of FePt Nanoparticles of Various Sizes for Magnetic Data Storage

    E-Print Network [OSTI]

    Laughlin, David E.

    to a modified reaction route based on Sun et al.[1] Hexane dispersions of nanoparticles were dried increases with particle size and with the temperature in the range 600 °C to 650 °C, being close to unity-assembly over large areas, and a narrow distribution of switching fields. The long storage time and high storage

  1. Distributed storage with communication costs

    E-Print Network [OSTI]

    Armstrong, Craig Kenneth

    2011-01-01T23:59:59.000Z

    5 Introduction to Coding for Distributed Storage The Repairflow graph for 1 repair with varying storage capac- itythe Capacity of Storage Nodes . . . 4.1 Characterizing

  2. Storage Space Request Aurora Facility

    E-Print Network [OSTI]

    Ickert-Bond, Steffi

    Storage Space Request Aurora Facility (1855 Marika) Department and Division: _______________________________________________________ Storage Contact: ____________________________________________________________ Name Phone and fax Fiscal Footage required: ______________ Brief Description of storage items

  3. Energy Storage | Argonne National Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Energy Storage Energy Storage The challenge of creating new advanced batteries and energy storage technologies is one of Argonne's key initiatives. By creating a multidisciplinary...

  4. Sandia National Laboratories: Energy Storage Multimedia Gallery

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    StorageEnergy Storage Multimedia Gallery Energy Storage Multimedia Gallery Images Videos Energy Storage Image Gallery Energy Storage B-Roll Videos Battery Abuse Testing Laboratory...

  5. Textile Drying Via Wood Gasification 

    E-Print Network [OSTI]

    McGowan, T. F.; Jape, A. D.

    1983-01-01T23:59:59.000Z

    This project was carried out to investigate the possibility of using wood gas as a direct replacement for natural gas in textile drying. The Georgia Tech updraft gasifier was used for the experimental program. During preliminary tests, the 1 million...

  6. Textile Drying Via Wood Gasification

    E-Print Network [OSTI]

    McGowan, T. F.; Jape, A. D.

    1983-01-01T23:59:59.000Z

    This project was carried out to investigate the possibility of using wood gas as a direct replacement for natural gas in textile drying. The Georgia Tech updraft gasifier was used for the experimental program. During preliminary tests, the 1 million...

  7. Model NOx storage systems: Storage capacity and thermal aging...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Model NOx storage systems: Storage capacity and thermal aging of BaOtheta- Al2O3NiAl(100). Model NOx storage systems: Storage capacity and thermal aging of BaOtheta- Al2O3...

  8. Storage Ring Operation Modes

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Longitudinal bunch profile and Up: APS Storage Ring Parameters Previous: Source Parameter Table Storage Ring Operation Modes Standard Operating Mode, top-up Fill pattern: 102 mA in...

  9. Underground Storage Tank Regulations

    Broader source: Energy.gov [DOE]

    The Underground Storage Tank Regulations is relevant to all energy projects that will require the use and building of pipelines, underground storage of any sorts, and/or electrical equipment. The...

  10. Cool Storage Performance 

    E-Print Network [OSTI]

    Eppelheimer, D. M.

    1985-01-01T23:59:59.000Z

    Utilities have promoted the use of electric heat and thermal storage to increase off peak usage of power. High daytime demand charges and enticing discounts for off peak power have been used as economic incentives to promote thermal storage systems...

  11. Safe Home Food Storage

    E-Print Network [OSTI]

    Van Laanen, Peggy

    2002-08-22T23:59:59.000Z

    Proper food storage can preserve food quality and prevent spoilage and food/borne illness. The specifics of pantry, refrigerator and freezer storage are given, along with helpful information on new packaging, label dates, etc. A comprehensive table...

  12. Geothermal Food Processors Agricultural Drying Low Temperature...

    Open Energy Info (EERE)

    Food Processors Agricultural Drying Low Temperature Geothermal Facility Jump to: navigation, search Name Geothermal Food Processors Agricultural Drying Low Temperature Geothermal...

  13. Energy Storage Systems

    SciTech Connect (OSTI)

    Conover, David R.

    2013-12-01T23:59:59.000Z

    Energy Storage Systems – An Old Idea Doing New Things with New Technology article for the International Assoication of ELectrical Inspectors

  14. UFD Storage and Transportation - Transportation Working Group Report

    SciTech Connect (OSTI)

    Maheras, Steven J.; Ross, Steven B.

    2011-08-01T23:59:59.000Z

    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 transportation importance was mirrored by the importance assigned by the UFD Storage Task. A few of the more significant differences are described in Section 3 of this report

  15. FOREST CENTRE STORAGE BUILDING

    E-Print Network [OSTI]

    deYoung, Brad

    FOREST CENTRE STORAGE BUILDING 3 4 5 6 7 8 UniversityDr. 2 1 G r e n f e l l D r i v e MULTI PURPOSE COURT STUDENT RESIDENCES GREEN HOUSE STUDENT RESIDENCES STUDENT RESIDENCES RECPLEX STORAGE BUILDING STORAGE BUILDING LIBRARY & COMPUTING FINE ARTS FOREST CENTRE ARTS &SCIENCE BUILDING ARTS &SCIENCE

  16. Vadose zone transport in dry forests of central Argentina: Role of land use

    E-Print Network [OSTI]

    Nacional de San Luis, Universidad

    on residual moisture flux approach (cumulative chloride versus cumulative water curves) suggested maximum of water below root zone, displacement of solutes, and rising water tables, affecting, in most extreme 78% to 99% of the chloride stock was leached, and total water storage was 30% higher than in the dry

  17. Sandia National Laboratories: Energy Storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    for Infrastructure Research and Innovation (CIRI), Concentrating Solar Power, Energy, Energy Storage, Energy Storage Systems, Facilities, Infrastructure Security, Materials...

  18. Groundwater and Terrestrial Water Storage

    E-Print Network [OSTI]

    Rodell, M; Chambers, D P; Famiglietti, J S

    2011-01-01T23:59:59.000Z

    T. E. Reilly, 2002: Flow and storage in groundwater systems.Estimating ground water storage changes in the Mississippistorage..

  19. Storage : DAS / SAN / NAS Dploiement

    E-Print Network [OSTI]

    Collette. Sébastien

    CH8 Divers Agenda · Storage : DAS / SAN / NAS · Déploiement · VLAN ­ 802.1Q · Gestion d · Sécurisation de Windows · Sécurisation de UNIX · Qu'est-ce que... ­ Firewall, VPN, IDS/IPS, PKI Storage : DAS, NAS, SAN #12;Storage : DAS, NAS, SAN · Direct Attached Storage · Network Attached Storage · Storage

  20. Storage Ring Revised March 1994

    E-Print Network [OSTI]

    Brookhaven National Laboratory - Experiment 821

    Chapter 8. Storage Ring Revised March 1994 8.1. Introduction -- 107 -- #12; 108 Storage Ring 8.2. Magnetic Design and Field Calculations 8.2.1. Conceptual Approach #12; Storage Ring 109 #12; 110 Storage Ring 8.2.2. Computer Aided Refined Pole Designs #12; Storage Ring 111 #12; 112 Storage Ring #12

  1. GenTegra DNA Tubes is another proprietary technology for stor-ing purified DNA in a `bone-dry', water-free environment. This new

    E-Print Network [OSTI]

    Cai, Long

    -dry', water-free environment. This new matrix protects DNA samples from hydrolysis and oxidation platforms, including Illumina and Affymetrix. Dry-state, room-temperature storage of DNA and RNA Water, Heather E McMahon & Michael E Hogan GenVault Corporation, 6190 Corte Del Cedro, Carlsbad, California 92011

  2. Cold Vacuum Drying (CVD) Facility General Service Helium System Design Description

    SciTech Connect (OSTI)

    SHAPLEY, B.J.

    2000-04-20T23:59:59.000Z

    The purpose of this System Design Description (SDD) is to describe the characteristics of the Cold Vacuum Drying (CVD) Facility general service helium system. The general service helium system is a general service facility process support system, but does include safety-class structures, systems and components (SSCs) providing protection to the offsite public. The general service helium system also performs safety-significant functions that provide protection to onsite workers. The general helium system essential function is to provide helium (He) to support process functions during all phases of facility operations. General service helium is used to purge the cask and the MCO in order to maintain their internal atmospheres below hydrogen flammability concentrations. The general service helium system also supplies helium to purge the process water conditioning (PWC) lines and components and the vacuum purge system (VPS) vacuum pump. The general service helium system, if available following an Safety Class Instrument and Control System (SCIC) Isolation and Purge (IS0 and PURGE) Trip, can provide an alternate general service helium system source to supply the Safety-Class Helium (SCHe) System.

  3. A Low-Tech, Low-Budget Storage Solution for High Level Radioactive Sources

    SciTech Connect (OSTI)

    Brett Carlsen; Ted Reed; Todd Johnson; John Weathersby; Joe Alexander; Dave Griffith; Douglas Hamelin

    2014-07-01T23:59:59.000Z

    The need for safe, secure, and economical storage of radioactive material becomes increasingly important as beneficial uses of radioactive material expand (increases inventory), as political instability rises (increases threat), and as final disposal and treatment facilities are delayed (increases inventory and storage duration). Several vendor-produced storage casks are available for this purpose but are often costly — due to the required design, analyses, and licensing costs. Thus the relatively high costs of currently accepted storage solutions may inhibit substantial improvements in safety and security that might otherwise be achieved. This is particularly true in areas of the world where the economic and/or the regulatory infrastructure may not provide the means and/or the justification for such an expense. This paper considers a relatively low-cost, low-technology radioactive material storage solution. The basic concept consists of a simple shielded storage container that can be fabricated locally using a steel pipe and a corrugated steel culvert as forms enclosing a concrete annulus. Benefits of such a system include 1) a low-tech solution that utilizes materials and skills available virtually anywhere in the world, 2) a readily scalable design that easily adapts to specific needs such as the geometry and radioactivity of the source term material), 3) flexible placement allows for free-standing above-ground or in-ground (i.e., below grade or bermed) installation, 4) the ability for future relocation without direct handling of sources, and 5) a long operational lifetime . ‘Le mieux est l’ennemi du bien’ (translated: The best is the enemy of good) applies to the management of radioactive materials – particularly where the economic and/or regulatory justification for additional investment is lacking. Development of a low-cost alternative that considerably enhances safety and security may lead to a greater overall risk reduction than insisting on solutions that remain economically and/or politically ‘out of reach’.

  4. EFFECT OF MECHANICAL CONDITIONING ON THIN-LAYER DRYING OF ENERGY SORGHUM (Sorghum bicolor (L.) Moench)

    SciTech Connect (OSTI)

    Ian J. Bonner; Kevin L. Kenney

    2012-10-01T23:59:59.000Z

    Cellulosic energy varieties of Sorghum bicolor (L.) Moench show promise as a bioenergy feedstock, however, high moisture content at the time of harvest results in unacceptable levels of degradation when stored in aerobic conditions. To safely store sorghum biomass for extended periods in baled format, the material must be dried to inhibit microbial growth. One possible solution is allowing the material to dry under natural in-field conditions. This study examines the differences in thin-layer drying rates of intact and conditioned sorghum under laboratory-controlled temperatures and relative humidity levels (20 degrees C and 30 degrees C from 40% to 85% relative humidity), and models experimental data using the Page’s Modified equation. The results demonstrate that conditioning drastically accelerates drying times. Relative humidity had a large impact on the time required to reach a safe storage moisture content for intact material (approximately 200 hours at 30 degrees C and 40% relative humidity and 400 hours at 30 degrees C and 70% relative humidity), but little to no impact on the thin-layer drying times of conditioned material (approximately 50 hours for all humidity levels < 70% at 30 degrees C). The drying equation parameters were influenced by temperature, relative humidity, initial moisture content, and material damage, allowing drying curves to be empirically predicted. The results of this study provide valuable information applicable to the agricultural community and to future research on drying simulation and management of energy sorghum.

  5. Heat storage duration

    SciTech Connect (OSTI)

    Balcomb, J.D.

    1981-01-01T23:59:59.000Z

    Both the amount and duration of heat storage in massive elements of a passive building are investigated. Data taken for one full winter in the Balcomb solar home are analyzed with the aid of sub-system simulation models. Heat storage duration is tallied into one-day intervals. Heat storage location is discussed and related to overall energy flows. The results are interpreted and conclusions drawn.

  6. Energy Storage Program Overview

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Barriers HEV & PHEV Technology Roadmaps R&D Timeline Overview 3 Develop electrochemical energy storage technologies which support the commercialization of hybrid and electric...

  7. Hydrogen Storage Related Links

    Broader source: Energy.gov [DOE]

    The following resources provide details about DOE-funded hydrogen storage activities, research plans and roadmaps, models and tools, and additional related links.

  8. Culex quinquefasciatus Storage Proteins

    E-Print Network [OSTI]

    2013-01-01T23:59:59.000Z

    and hemolymph proteins of Cx. quinquefasciatus . A and B:of typical storage proteins in Cx. quinquefasciatus.Fourth-instar Cx. quinquefasciatus larvae and early pupae

  9. HEATS: Thermal Energy Storage

    SciTech Connect (OSTI)

    None

    2012-01-01T23:59:59.000Z

    HEATS Project: The 15 projects that make up ARPA-E’s HEATS program, short for “High Energy Advanced Thermal Storage,” seek to develop revolutionary, cost-effective ways to store thermal energy. HEATS focuses on 3 specific areas: 1) developing high-temperature solar thermal energy storage capable of cost-effectively delivering electricity around the clock and thermal energy storage for nuclear power plants capable of cost-effectively meeting peak demand, 2) creating synthetic fuel efficiently from sunlight by converting sunlight into heat, and 3) using thermal energy storage to improve the driving range of electric vehicles (EVs) and also enable thermal management of internal combustion engine vehicles.

  10. SUPERCONDUCTING MAGNETIC ENERGY STORAGE

    E-Print Network [OSTI]

    Hassenzahl, W.

    2011-01-01T23:59:59.000Z

    and R. W . BOOIll, "Superconductive Energy Storage Inducand H. A. Peterson, "Superconductive E nergy S torage forMeeting, Janua ry N. Mohan, "Superconductive Energy S torage

  11. SUPERCONDUCTING MAGNETIC ENERGY STORAGE

    E-Print Network [OSTI]

    Hassenzahl, W.

    2011-01-01T23:59:59.000Z

    Design of the BPA Superconducting 30-MJ Energy Storagefor a Utility Scale Superconducting Magnetic Energy Storagefor a Lnrge Scale Superconducting Magnetic Energy Storage

  12. Evaluation of Storage for Transportation Equipment, Unfueled Convertors, and Fueled Convertors at the INL for the Radioisotope Power Systems Program

    SciTech Connect (OSTI)

    S. G. Johnson; K. L. Lively

    2010-05-01T23:59:59.000Z

    This report contains an evaluation of the storage conditions required for several key components and/or systems of the Radioisotope Power Systems (RPS) Program at the Idaho National Laboratory (INL). These components/systems (transportation equipment, i.e., type ‘B’ shipping casks and the radioisotope thermo-electric generator transportation systems (RTGTS), the unfueled convertors, i.e., multi-hundred watt (MHW) and general purpose heat source (GPHS) RTGs, and fueled convertors of several types) are currently stored in several facilities at the Materials and Fuels Complex (MFC) site. For various reasons related to competing missions, inherent growth of the RPS mission at the INL and enhanced efficiency, it is necessary to evaluate their current storage situation and recommend the approach that should be pursued going forward for storage of these vital RPS components and systems. The reasons that drive this evaluation include, but are not limited to the following: 1) conflict with other missions at the INL of higher priority, 2) increasing demands from the INL RPS Program that exceed the physical capacity of the current storage areas and 3) the ability to enhance our current capability to care for our equipment, decrease maintenance costs and increase the readiness posture of the systems.

  13. Secondary Storage Management Himanshu Gupta

    E-Print Network [OSTI]

    Gupta, Himanshu

    Secondary Storage Management Himanshu Gupta Storage­1 #12;Outline · Memory Hierarchy · Disk Records/Fields · Deletions and Insertions of Records Himanshu Gupta Storage­2 #12;Himanshu Gupta Storage­3 Memory Hierarchy Cache (1 MB; 1-5 nsec) Main Memory (GBs; 10-100 nsec) Secondary Storage

  14. Optimal Storage Allocation for Serial

    E-Print Network [OSTI]

    Yechiali, Uri

    Optimal Storage Allocation for Serial Haim Mendelson, Joseph S. Pliskin, and Uri Yechiali Tel Aviv reside on a direct-access storage device in which storage space is limited. Records are added allocating storage space to the files. Key Words and Phrases: serial files, storage allocation

  15. Sandia National Laboratories: implement energy storage projects

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    implement energy storage projects Sandian Spoke at the New York Energy Storage Expo On December 12, 2014, in Energy, Energy Storage, Energy Storage Systems, Grid Integration,...

  16. Sandia National Laboratories: Stationary Energy Storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    StorageStationary Energy Storage Stationary Energy Storage The 1 MW Energy Storage Test Pad integrated with renewable energy generation at Sandia's Distributed Energy Technology...

  17. Sandia National Laboratories: Batteries & Energy Storage Publications

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    StorageBatteries & Energy Storage Publications Batteries & Energy Storage Publications Batteries & Energy Storage Fact Sheets Achieving Higher Energy Density in Flow Batteries at...

  18. Sandia National Laboratories: evaluate energy storage opportunity

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    energy storage opportunity Sandian Spoke at the New York Energy Storage Expo On December 12, 2014, in Energy, Energy Storage, Energy Storage Systems, Grid Integration,...

  19. THERMAL ENERGY STORAGE IN AQUIFERS WORKSHOP

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    The Legalization of Ground Water Storage," Water Resourcesprocedure to above ground storage of heat in huge insulatedthis project is heat storage in ground-water regions storage

  20. Sandia Energy - Energy Storage Test Pad (ESTP)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Storage Test Pad (ESTP) Home Energy Permalink Gallery Evaluating Powerful Batteries for Modular Electric Grid Energy Storage Energy, Energy Storage, Energy Storage Systems, Energy...

  1. Energy storage capacitors

    SciTech Connect (OSTI)

    Sarjeant, W.J.

    1984-01-01T23:59:59.000Z

    The properties of capacitors are reviewed in general, including dielectrics, induced polarization, and permanent polarization. Then capacitance characteristics are discussed and modelled. These include temperature range, voltage, equivalent series resistance, capacitive reactance, impedance, dissipation factor, humidity and frequency effects, storage temperature and time, and lifetime. Applications of energy storage capacitors are then discussed. (LEW)

  2. Nonlinear Dynamics of Dry Friction

    E-Print Network [OSTI]

    Franz-Josef Elmer

    1997-07-01T23:59:59.000Z

    The dynamical behavior caused by dry friction is studied for a spring-block system pulled with constant velocity over a surface. The dynamical consequences of a general type of phenomenological friction law (stick-time dependent static friction, velocity dependent kinetic friction) are investigated. Three types of motion are possible: Stick-slip motion, continuous sliding, and oscillations without sticking events. A rather complete discussion of local and global bifurcation scenarios of these attractors and their unstable counterparts is present.

  3. Gas Storage Technology Consortium

    SciTech Connect (OSTI)

    Joel L. Morrison; Sharon L. Elder

    2006-07-06T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission & distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services, and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is critical in meeting the needs of these new markets. In order to address the gas storage needs of the natural gas industry, an industry-driven consortium was created--the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance operational flexibility and deliverability of the Nation's gas storage system, and provide a cost effective, safe, and reliable supply of natural gas to meet domestic demand. This report addresses the activities for the quarterly period of April 1 to June 30, 2006. Key activities during this time period include: (1) Develop and process subcontract agreements for the eight projects selected for cofunding at the February 2006 GSTC Meeting; (2) Compiling and distributing the three 2004 project final reports to the GSTC Full members; (3) Develop template, compile listserv, and draft first GSTC Insider online newsletter; (4) Continue membership recruitment; (5) Identify projects and finalize agenda for the fall GSTC/AGA Underground Storage Committee Technology Transfer Workshop in San Francisco, CA; and (6) Identify projects and prepare draft agenda for the fall GSTC Technology Transfer Workshop in Pittsburgh, PA.

  4. Drying and Storing Cooperative Extension Service

    E-Print Network [OSTI]

    Mukhtar, Saqib

    . Sunflowers Joseph P. Harner Extension Agriculture Engineer The fire hazard is DECREASED when the fan can draw for attachment to the drying fan. Guidelines for drying sunflowers are: 1. 2. 3. 4. Use good housekeeping practices. Clean up around the dryer and in the plenum chamber daily. Do not over dry. Ensure continuous

  5. Uranium Ore Uranium is extracted

    E-Print Network [OSTI]

    be discharged to water. Radioactive Wastes--Wastes managed for their radioactive content. Spent Nuclear Fuels--Fuel plants with reactors that use water for moderating nuclear reactions and cooling. Spent Nuclear Fuel Used or"spent"nuclear fuel is stored in pools, or in specially designed dry storage casks. Fabrication

  6. Sandia National Laboratories: Energy Storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Energy Storage Electric Car Challenge Sparks Students' STEM Interest On January 9, 2015, in Energy, Energy Storage, News, News & Events, Partnership, Transportation Energy Aspiring...

  7. Improving energy storage devices | EMSL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    energy storage devices Improving energy storage devices Released: April 15, 2014 Lithium-sulfur batteries last longer with nanomaterial-packed cathode A new PNNL-developed...

  8. Sandia National Laboratories: Energy Storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Capture & Storage, Center for Infrastructure Research and Innovation (CIRI), Energy, Energy Storage, Facilities, Livermore Valley Open Campus (LVOC), Materials Science, News,...

  9. Sandia National Laboratories: Energy Storage

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Collaboration On May 28, 2014, in Biofuels, CRF, Distribution Grid Integration, Energy, Energy Storage, Energy Storage Systems, Energy Surety, Facilities, Grid Integration,...

  10. Ultrafine hydrogen storage powders

    DOE Patents [OSTI]

    Anderson, Iver E. (Ames, IA); Ellis, Timothy W. (Doylestown, PA); Pecharsky, Vitalij K. (Ames, IA); Ting, Jason (Ames, IA); Terpstra, Robert (Ames, IA); Bowman, Robert C. (La Mesa, CA); Witham, Charles K. (Pasadena, CA); Fultz, Brent T. (Pasadena, CA); Bugga, Ratnakumar V. (Arcadia, CA)

    2000-06-13T23:59:59.000Z

    A method of making hydrogen storage powder resistant to fracture in service involves forming a melt having the appropriate composition for the hydrogen storage material, such, for example, LaNi.sub.5 and other AB.sub.5 type materials and AB.sub.5+x materials, where x is from about -2.5 to about +2.5, including x=0, and the melt is gas atomized under conditions of melt temperature and atomizing gas pressure to form generally spherical powder particles. The hydrogen storage powder exhibits improved chemcial homogeneity as a result of rapid solidfication from the melt and small particle size that is more resistant to microcracking during hydrogen absorption/desorption cycling. A hydrogen storage component, such as an electrode for a battery or electrochemical fuel cell, made from the gas atomized hydrogen storage material is resistant to hydrogen degradation upon hydrogen absorption/desorption that occurs for example, during charging/discharging of a battery. Such hydrogen storage components can be made by consolidating and optionally sintering the gas atomized hydrogen storage powder or alternately by shaping the gas atomized powder and a suitable binder to a desired configuration in a mold or die.

  11. Gas Storage Technology Consortium

    SciTech Connect (OSTI)

    Joel L. Morrison; Sharon L. Elder

    2007-06-30T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is crucial in meeting the needs of these new markets. To address the gas storage needs of the natural gas industry, an industry-driven consortium was created--the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance the operational flexibility and deliverability of the nation's gas storage system, and provide a cost-effective, safe, and reliable supply of natural gas to meet domestic demand. This report addresses the activities for the quarterly period of April 1, 2007 through June 30, 2007. Key activities during this time period included: (1) Organizing and hosting the 2007 GSTC Spring Meeting; (2) Identifying the 2007 GSTC projects, issuing award or declination letters, and begin drafting subcontracts; (3) 2007 project mentoring teams identified; (4) New NETL Project Manager; (5) Preliminary planning for the 2007 GSTC Fall Meeting; (6) Collecting and compiling the 2005 GSTC project final reports; and (7) Outreach and communications.

  12. Gas Storage Technology Consortium

    SciTech Connect (OSTI)

    Joel L. Morrison; Sharon L. Elder

    2006-05-10T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services, and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is critical in meeting the needs of these new markets. In order to address the gas storage needs of the natural gas industry, an industry-driven consortium was created--the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance operational flexibility and deliverability of the Nation's gas storage system, and provide a cost effective, safe, and reliable supply of natural gas to meet domestic demand. This report addresses the activities for the quarterly period of January 1, 2006 through March 31, 2006. Activities during this time period were: (1) Organize and host the 2006 Spring Meeting in San Diego, CA on February 21-22, 2006; (2) Award 8 projects for co-funding by GSTC for 2006; (3) New members recruitment; and (4) Improving communications.

  13. Gas Storage Technology Consortium

    SciTech Connect (OSTI)

    Joel L. Morrison; Sharon L. Elder

    2007-03-31T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is crucial in meeting the needs of these new markets. To address the gas storage needs of the natural gas industry, an industry-driven consortium was created - the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance the operational flexibility and deliverability of the nation's gas storage system, and provide a cost-effective, safe, and reliable supply of natural gas to meet domestic demand. This report addresses the activities for the quarterly period of January1, 2007 through March 31, 2007. Key activities during this time period included: {lg_bullet} Drafting and distributing the 2007 RFP; {lg_bullet} Identifying and securing a meeting site for the GSTC 2007 Spring Proposal Meeting; {lg_bullet} Scheduling and participating in two (2) project mentoring conference calls; {lg_bullet} Conducting elections for four Executive Council seats; {lg_bullet} Collecting and compiling the 2005 GSTC Final Project Reports; and {lg_bullet} Outreach and communications.

  14. Gas Storage Technology Consortium

    SciTech Connect (OSTI)

    Joel Morrison

    2005-09-14T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services, and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is critical in meeting the needs of these new markets. In order to address the gas storage needs of the natural gas industry, an industry driven consortium was created--the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance operational flexibility and deliverability of the Nation's gas storage system, and provide a cost effective, safe, and reliable supply of natural gas to meet domestic demand. This report addresses the activities for the quarterly period of April 1, 2005 through June 30, 2005. During this time period efforts were directed toward (1) GSTC administration changes, (2) participating in the American Gas Association Operations Conference and Biennial Exhibition, (3) issuing a Request for Proposals (RFP) for proposal solicitation for funding, and (4) organizing the proposal selection meeting.

  15. Hydrogen storage composition and method

    DOE Patents [OSTI]

    Wicks, G.G.; Heung, L.K.

    1994-01-01T23:59:59.000Z

    A hydrogen storage composition based on a metal hydride dispersed in an aerogel prepared by a sol-gel process. The starting material for the aerogel is an organometallic compound, including the alkoxysilanes, organometals of the form M(OR){sub X} where R is an organic ligand of the form C{sub n}H{sub 2n+1}, and organometals of the form MO{sub x}Ry where R is an alkyl group, where M is an oxide-forming metal, n, x and y are integers and y is two less than the valence of M. A sol is prepared by combining the starting material, alcohol, water, and an acid. The sol is conditioned to the proper viscosity and a hydride in the form of a fine powder is added. The mixture is polymerized and dried under supercritical conditions. The final product is a composition having a hydride uniformly dispersed throughout an inert, stable and highly porous matrix. It is capable of absorbing up to 30 motes of hydrogen per kilogram at room temperature and pressure, rapidly and reversibly. Hydrogen absorbed by the composition can be readily be recovered by heat or evacuation.

  16. Hydrogen storage composition and method

    DOE Patents [OSTI]

    Heung, Leung K (Aiken, SC); Wicks, George G. (Aiken, SC)

    2003-01-01T23:59:59.000Z

    A hydrogen storage composition based on a metal hydride dispersed in an aerogel prepared by a sol-gel process. The starting material for the aerogel is an organometallic compound, including the alkoxysilanes, organometals of the form M(OR)x and MOxRy, where R is an alkyl group of the form C.sub.n H.sub.2n+1, M is an oxide-forming metal, n, x, and y are integers, and y is two less than the valence of M. A sol is prepared by combining the starting material, alcohol, water, and an acid. The sol is conditioned to the proper viscosity and a hydride in the form of a fine powder is added. The mixture is polymerized and dried under supercritical conditions. The final product is a composition having a hydride uniformly dispersed throughout an inert, stable and highly porous matrix. It is capable of absorbing up to 30 moles of hydrogen per kilogram at room temperature and pressure, rapidly and reversibly. Hydrogen absorbed by the composition can be readily be recovered by heat or evacuation.

  17. Influence of Airflow on Laboratory Storage of High Moisture Corn Stover

    SciTech Connect (OSTI)

    Lynn M. Wendt; Ian J. Bonner; Amber N. Hoover; Rachel M. Emerson; William A. Smith

    2014-04-01T23:59:59.000Z

    Storing high moisture biomass for bioenergy use is a reality in many areas of the country where wet harvest conditions and environmental factors prevent dry storage from being feasible. Aerobic storage of high moisture biomass leads to microbial degradation and self-heating, but oxygen limitation can aid in material preservation. To understand the influence of oxygen presence on high moisture biomass (50 %, wet basis), three airflow rates were tested on corn stover stored in laboratory reactors. Temperature, carbon dioxide production, dry matter loss, chemical composition, fungal abundance, pH, and organic acids were used to monitor the effects of airflow on storage conditions. The results of this work indicate that oxygen availability impacts both the duration of self-heating and the severity of dry matter loss. High airflow systems experienced the greatest initial rates of loss but a shortened microbially active period that limited total dry matter loss (19 %). Intermediate airflow had improved preservation in short-term storage compared to high airflow systems but accumulated the greatest dry matter loss over time (up to 27 %) as a result of an extended microbially active period. Low airflow systems displayed the best performance with the lowest rates of loss and total loss (10 %) in storage at 50 days. Total structural sugar levels of the stored material were preserved, although glucan enrichment and xylan loss were documented in the high and intermediate flow conditions. By understanding the role of oxygen availability on biomass storage performance, the requirements for high moisture storage solutions may begin to be experimentally defined.

  18. Method of preparing nuclear wastes for tansportation and interim storage

    DOE Patents [OSTI]

    Bandyopadhyay, Gautam (Naperville, IL); Galvin, Thomas M. (Darien, IL)

    1984-01-01T23:59:59.000Z

    Nuclear waste is formed into a substantially water-insoluble solid for temporary storage and transportation by mixing the calcined waste with at least 10 weight percent powdered anhydrous sodium silicate to form a mixture and subjecting the mixture to a high humidity environment for a period of time sufficient to form cementitious bonds by chemical reaction. The method is suitable for preparing an interim waste form from dried high level radioactive wastes.

  19. Storage Exchange: A Global Trading Platform for Storage Services

    E-Print Network [OSTI]

    Melbourne, University of

    Storage Exchange: A Global Trading Platform for Storage Services Martin Placek and Rajkumar Buyya,raj}@csse.unimelb.edu.au Abstract. The Storage Exchange (SX) is a new platform allowing stor- age to be treated as a tradeable resource. Organisations with varying storage requirements can use the SX platform to trade and exchange

  20. Building Trust in Storage Outsourcing: Secure Accounting of Utility Storage

    E-Print Network [OSTI]

    Minnesota, University of

    Building Trust in Storage Outsourcing: Secure Accounting of Utility Storage Vishal Kher Yongdae Kim are witnessing a revival of Storage Service Providers (SSP) in the form of new vendors as well as traditional players. While storage outsourcing is cost-effective, many companies are hesitating to outsource

  1. Technical basis for extending storage of the UK's advanced gas-cooled reactor fuel

    SciTech Connect (OSTI)

    Hambley, D.I. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

    2013-07-01T23:59:59.000Z

    The UK Nuclear Decommissioning Agency has recently declared a date for cessation of reprocessing of oxide fuel from the UK's Advanced Gas-cooled Reactors (AGRs). This will fundamentally change the management of AGR fuel: from short term storage followed by reprocessing to long term fuel storage followed, in all likelihood, by geological disposal. In terms of infrastructure, the UK has an existing, modern wet storage asset that can be adapted for centralised long term storage of dismantled AGR fuel under the required pond water chemistry. No AGR dry stores exist, although small quantities of fuel have been stored dry as part of experimental programmes in the past. These experimental programmes have shown concerns about corrosion rates.

  2. Microwave drying of ferric oxide pellets

    SciTech Connect (OSTI)

    Pickles, C.A.; Xia, D.K. [Queens` Univ., Kingston, Ontario (Canada). Dept. of Materials and Metallurgical Engineering

    1997-12-31T23:59:59.000Z

    The application of microwave energy for the drying of ferric oxide pellets has been investigated and evaluated. It is shown that the microwave drying rates are much higher than those observed in the conventional process. Also there is some potential for improved quality of the product. As a stand-alone technology it is unlikely that microwave drying would be economical for pellets due to the low cost of conventional fuels. However, based on an understanding of the drying mechanisms in the conventional process and in the microwave process, it is shown that microwave-assisted drying offers considerable potential. In this hybrid process, the advantages of the two drying techniques are combined to provide an improved drying process.

  3. Dry-cleaning of graphene

    SciTech Connect (OSTI)

    Algara-Siller, Gerardo [Central Facility for Electron Microscopy, Group of Electron Microscopy of Materials Science, Ulm University, Albert-Einstein-Allee 11, Ulm 89081 (Germany); Department of Chemistry, Technical University Ilmenau, Weimarer Strasse 25, Ilmenau 98693 (Germany); Lehtinen, Ossi; Kaiser, Ute, E-mail: ute.kaiser@uni-ulm.de [Central Facility for Electron Microscopy, Group of Electron Microscopy of Materials Science, Ulm University, Albert-Einstein-Allee 11, Ulm 89081 (Germany); Turchanin, Andrey [Faculty of Physics, University of Bielefeld, Universitätsstr. 25, Bielefeld 33615 (Germany)

    2014-04-14T23:59:59.000Z

    Studies of the structural and electronic properties of graphene in its pristine state are hindered by hydrocarbon contamination on the surfaces. Also, in many applications, contamination reduces the performance of graphene. Contamination is introduced during sample preparation and is adsorbed also directly from air. Here, we report on the development of a simple dry-cleaning method for producing large atomically clean areas in free-standing graphene. The cleanness of graphene is proven using aberration-corrected high-resolution transmission electron microscopy and electron spectroscopy.

  4. Vapor Transport in Dry Soils

    SciTech Connect (OSTI)

    Gee, Glendon W.; Ward, Anderson L.

    2001-11-16T23:59:59.000Z

    Water-vapor movement in soils is a complex process, controlled by both diffusion and advection and influenced by pressure and thermal gradients acting across tortuous flow paths. Wide-ranging interest in water-vapor transport includes both theoretical and practical aspects. Just how pressure and thermal gradients enhance water-vapor flow is still not completely understood and subject to ongoing research. Practical aspects include dryland farming (surface mulching), water harvesting (aerial wells), fertilizer placement, and migration of contaminants at waste-sites. The following article describes the processes and practical applications of water-vapor transport, with emphasis on unsaturated (dry) soil systems.

  5. SUPERCONDUCTING MAGNETIC ENERGY STORAGE

    E-Print Network [OSTI]

    Hassenzahl, W.

    2011-01-01T23:59:59.000Z

    Encrgy Storage Plant" , EPRI Report EM-3457, April 1984. [4521st century. REFERENCES The EPRI Regional Systems preparedby J. J. Mulvaney, EPRI Report EPRI P-19S0SR, (1981). [2J O.

  6. Marketing Cool Storage Technology 

    E-Print Network [OSTI]

    McCannon, L.

    1987-01-01T23:59:59.000Z

    in the field. The International Thermal Storage Advisory Council was formed to help meet this perceived need. This paper will review activities of EPRI and ITSAC to achieve widespread acceptance of the technology....

  7. Hydrogen storage compositions

    SciTech Connect (OSTI)

    Li, Wen; Vajo, John J.; Cumberland, Robert W.; Liu, Ping

    2011-04-19T23:59:59.000Z

    Compositions for hydrogen storage and methods of making such compositions employ an alloy that exhibits reversible formation/deformation of BH4- anions. The composition includes a ternary alloy including magnesium, boron and a metal and a metal hydride. The ternary alloy and the metal hydride are present in an amount sufficient to render the composition capable of hydrogen storage. The molar ratio of the metal to magnesium and boron in the alloy is such that the alloy exhibits reversible formation/deformation of BH4- anions. The hydrogen storage composition is prepared by combining magnesium, boron and a metal to prepare a ternary alloy and combining the ternary alloy with a metal hydride to form the hydrogen storage composition.

  8. Hydrogen storage compositions

    DOE Patents [OSTI]

    Li, Wen; Vajo, John J.; Cumberland, Robert W.; Liu, Ping

    2011-04-19T23:59:59.000Z

    Compositions for hydrogen storage and methods of making such compositions employ an alloy that exhibits reversible formation/deformation of BH.sub.4.sup.- anions. The composition includes a ternary alloy including magnesium, boron and a metal and a metal hydride. The ternary alloy and the metal hydride are present in an amount sufficient to render the composition capable of hydrogen storage. The molar ratio of the metal to magnesium and boron in the alloy is such that the alloy exhibits reversible formation/deformation of BH.sub.4.sup.- anions. The hydrogen storage composition is prepared by combining magnesium, boron and a metal to prepare a ternary alloy and combining the ternary alloy with a metal hydride to form the hydrogen storage composition.

  9. APS Storage Ring Parameters

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Main Parameters APS Storage Ring Parameters M. Borland, G. Decker, L. Emery, W. Guo, K. Harkay, V. Sajaev, C.-Y. Yao Advanced Photon Source September 8, 2010 This document list the...

  10. Stasis: Flexible Transactional Storage

    E-Print Network [OSTI]

    Sears, Russell C.

    2009-01-01T23:59:59.000Z

    He and Bowei Du implemented Oasys, and helped with my firstwas built on top of a C++ object persistence library, Oasys.Oasys uses plug-in storage modules that implement persistent

  11. Gas Storage Act (Illinois)

    Broader source: Energy.gov [DOE]

    Any corporation which is engaged in or desires to engage in, the distribution, transportation or storage of natural gas or manufactured gas, which gas, in whole or in part, is intended for ultimate...

  12. SUPERCONDUCTING MAGNETIC ENERGY STORAGE

    E-Print Network [OSTI]

    Hassenzahl, W.

    2011-01-01T23:59:59.000Z

    World's First 290 MW Gas Turbine Air Storage Peaking Plant",hydro e lectric plants and gas turbines, are less effectedelectricity. For a gas turbine the conversion efficiency may

  13. Storage Tanks (Arkansas)

    Broader source: Energy.gov [DOE]

    The Storage Tanks regulations is a set of rules and permit requirements mandated by the Arkansas Pollution and Ecology Commission in order to protect the public health and the lands and the waters...

  14. Thermal Energy Storage

    SciTech Connect (OSTI)

    Rutberg, Michael; Hastbacka, Mildred; Cooperman, Alissa; Bouza, Antonio

    2013-06-05T23:59:59.000Z

    The article discusses thermal energy storage technologies. This article addresses benefits of TES at both the building site and the electricity generation source. The energy savings and market potential of thermal energy store are reviewed as well.

  15. Energy Storage 101

    Broader source: Energy.gov (indexed) [DOE]

    the storage of heat or cold between opposing seasons in deep aquifers or bedrock. A wind-up clock stores potential energy, in this case mechanical, in the spring tension. ...

  16. Storage management solutions Buyer's guide: purchasing criteria

    E-Print Network [OSTI]

    Storage management solutions Buyer's guide: purchasing criteria Manage your storage to meet service storage environment cohesively As new guidelines or regulations surface, storage administrators receive increasing numbers of requests for change (RFCs) in storage provisioning. Simultaneously, routine changes

  17. The effects of storage on starch characteristics and chip quality of Texas Irish potatoes

    E-Print Network [OSTI]

    McDonald, Roy E

    1966-01-01T23:59:59.000Z

    of Department c Member) (Member) May, 1966 TABLE OF CONTENTS Chapter I. INTRODUCTION II. REVIEW OF LITERATURE Page Starch Starch Grain Size Starch Density Effect of Storage on Specific Gravity Effect of Storage on Starch Potato Chips III.... MATERIALS AND METHODS Measurement of Chemical and Physical Changes Specific Gravity Starch Grain Size Dry Matter Percentage Starch Density Starch Percent Chip Quality Chip Qo1or'~ Chip Oil Content Relationship of Starch Characteristics and Chip...

  18. Storage In C Matt Bishop

    E-Print Network [OSTI]

    Bishop, Matt

    Storage In C Matt Bishop Research Institute for Advanced Computer Science NASA Ames Research Center. Intimately bound with the idea of scope is that of storage. When a program defines a variable, the compiler storage (such as on a stack) or as more permanent storage (in data space.) Recall that the format of a C

  19. Storage In C Matt Bishop

    E-Print Network [OSTI]

    Bishop, Matt

    Storage In C Matt Bishop Research Institute for Advanced Computer Science NASA Ames Research Center. Intimately bound with the idea of scope is that of storage. When a program deÞnes a variable, the compiler storage (such as on a stack) or as more permanent storage (in data space.) Recall that the format of a C

  20. Dry melting of high albite

    SciTech Connect (OSTI)

    Anovitz, L.M.: Blencoe, J.G.

    1999-12-01T23:59:59.000Z

    The properties of albitic melts are central to thermodynamic models for synthetic and natural granitic liquids. The authors have analyzed published phase-equilibrium and thermodynamic data for the dry fusion of high albite to develop a more accurate equation for the Biggs free energy of this reaction to 30 kbar and 1,400 C. Strict criteria for reaction reversal were sued to evaluate the phase-equilibrium data, and the thermodynamic properties of solid and liquid albite were evaluated using the published uncertainties in the original measurements. Results suggest that neither available phase-equilibrium experiments nor thermodynamic data tightly constrain the location of the reaction. Experimental solidus temperatures at 1 atm range from 1,100 to 1,120 C. High-pressure experiments were not reversed completely and may have been affected by several sources of error, but the apparent inconsistencies among the results of the various experimentalists are eliminated when only half-reversal data are considered. Uncertainties in thermodynamic data yield large variations in permissible reaction slopes. Disparities between experimental and calculated melting curves are, therefore, largely attributable to these difficulties, and there is no fundamental disagreement between the available phase-equilibrium and thermodynamic data for the dry melting of albite. Consequently, complex speciation models for albitic melts, based on the assumption that these discrepancies represent a real characteristic of the system, are unjustified at this time.

  1. DOE Global Energy Storage Database

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    The DOE International Energy Storage Database has more than 400 documented energy storage projects from 34 countries around the world. The database provides free, up-to-date information on grid-connected energy storage projects and relevant state and federal policies. More than 50 energy storage technologies are represented worldwide, including multiple battery technologies, compressed air energy storage, flywheels, gravel energy storage, hydrogen energy storage, pumped hydroelectric, superconducting magnetic energy storage, and thermal energy storage. The policy section of the database shows 18 federal and state policies addressing grid-connected energy storage, from rules and regulations to tariffs and other financial incentives. It is funded through DOE’s Sandia National Laboratories, and has been operating since January 2012.

  2. Gas Storage Technology Consortium

    SciTech Connect (OSTI)

    Joel L. Morrison; Sharon L. Elder

    2006-09-30T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services, and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is critical in meeting the needs of these new markets. In order to address the gas storage needs of the natural gas industry, an industry-driven consortium was created-the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance operational flexibility and deliverability of the Nation's gas storage system, and provide a cost effective, safe, and reliable supply of natural gas to meet domestic demand. This report addresses the activities for the quarterly period of July 1, 2006 to September 30, 2006. Key activities during this time period include: {lg_bullet} Subaward contracts for all 2006 GSTC projects completed; {lg_bullet} Implement a formal project mentoring process by a mentor team; {lg_bullet} Upcoming Technology Transfer meetings: {sm_bullet} Finalize agenda for the American Gas Association Fall Underground Storage Committee/GSTC Technology Transfer Meeting in San Francisco, CA. on October 4, 2006; {sm_bullet} Identify projects and finalize agenda for the Fall GSTC Technology Transfer Meeting, Pittsburgh, PA on November 8, 2006; {lg_bullet} Draft and compile an electronic newsletter, the GSTC Insider; and {lg_bullet} New members update.

  3. GAS STORAGE TECHNOLOGY CONSORTIUM

    SciTech Connect (OSTI)

    Robert W. Watson

    2004-04-17T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services, and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is critical in meeting the needs of these new markets. In order to address the gas storage needs of the natural gas industry, an industry-driven consortium was created--the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance operational flexibility and deliverability of the Nation's gas storage system, and provide a cost effective, safe, and reliable supply of natural gas to meet domestic demand. To accomplish this objective, the project is divided into three phases that are managed and directed by the GSTC Coordinator. Base funding for the consortium is provided by the U.S. Department of Energy (DOE). In addition, funding is anticipated from the Gas Technology Institute (GTI). The first phase, Phase 1A, was initiated on September 30, 2003, and is scheduled for completion on March 31, 2004. Phase 1A of the project includes the creation of the GSTC structure, development of constitution (by-laws) for the consortium, and development and refinement of a technical approach (work plan) for deliverability enhancement and reservoir management. This report deals with the second 3-months of the project and encompasses the period December 31, 2003, through March 31, 2003. During this 3-month, the dialogue of individuals representing the storage industry, universities and the Department of energy was continued and resulted in a constitution for the operation of the consortium and a draft of the initial Request for Proposals (RFP).

  4. GAS STORAGE TECHNOLOGY CONSORTIUM

    SciTech Connect (OSTI)

    Robert W. Watson

    2004-10-18T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services, and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is critical in meeting the needs of these new markets. In order to address the gas storage needs of the natural gas industry, an industry-driven consortium was created--the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance operational flexibility and deliverability of the Nation's gas storage system, and provide a cost effective, safe, and reliable supply of natural gas to meet domestic demand. To accomplish this objective, the project is divided into three phases that are managed and directed by the GSTC Coordinator. The first phase, Phase 1A, was initiated on September 30, 2003, and was completed on March 31, 2004. Phase 1A of the project included the creation of the GSTC structure, development and refinement of a technical approach (work plan) for deliverability enhancement and reservoir management. This report deals with Phase 1B and encompasses the period July 1, 2004, through September 30, 2004. During this time period there were three main activities. First was the ongoing negotiations of the four sub-awards working toward signed contracts with the various organizations involved. Second, an Executive Council meeting was held at Penn State September 9, 2004. And third, the GSTC participated in the SPE Eastern Regional Meeting in Charleston, West Virginia, on September 16th and 17th. We hosted a display booth with the Stripper Well Consortium.

  5. GAS STORAGE TECHNOLOGY CONSORTIUM

    SciTech Connect (OSTI)

    Robert W. Watson

    2004-07-15T23:59:59.000Z

    Gas storage is a critical element in the natural gas industry. Producers, transmission and distribution companies, marketers, and end users all benefit directly from the load balancing function of storage. The unbundling process has fundamentally changed the way storage is used and valued. As an unbundled service, the value of storage is being recovered at rates that reflect its value. Moreover, the marketplace has differentiated between various types of storage services, and has increasingly rewarded flexibility, safety, and reliability. The size of the natural gas market has increased and is projected to continue to increase towards 30 trillion cubic feet (TCF) over the next 10 to 15 years. Much of this increase is projected to come from electric generation, particularly peaking units. Gas storage, particularly the flexible services that are most suited to electric loads, is critical in meeting the needs of these new markets. In order to address the gas storage needs of the natural gas industry, an industry-driven consortium was created--the Gas Storage Technology Consortium (GSTC). The objective of the GSTC is to provide a means to accomplish industry-driven research and development designed to enhance operational flexibility and deliverability of the Nation's gas storage system, and provide a cost effective, safe, and reliable supply of natural gas to meet domestic demand. To accomplish this objective, the project is divided into three phases that are managed and directed by the GSTC Coordinator. Base funding for the consortium is provided by the U.S. Department of Energy (DOE). In addition, funding is anticipated from the Gas Technology Institute (GTI). The first phase, Phase 1A, was initiated on September 30, 2003, and was completed on March 31, 2004. Phase 1A of the project included the creation of the GSTC structure, development and refinement of a technical approach (work plan) for deliverability enhancement and reservoir management. This report deals with Phase 1B and encompasses the period April 1, 2004, through June 30, 2004. During this 3-month period, a Request for Proposals (RFP) was made. A total of 17 proposals were submitted to the GSTC. A proposal selection meeting was held June 9-10, 2004 in Morgantown, West Virginia. Of the 17 proposals, 6 were selected for funding.

  6. Freeze-drying bovine spermatozoa

    E-Print Network [OSTI]

    Faris, Harvey Lee

    1965-01-01T23:59:59.000Z

    ~~to t~ roi'ipxg QQ ca dry ai gjuu QQjQigog aud ta Qst~~co cho ~~grso Qg 86lhVdratiea KXpkos Q~Kd Wlthstsud?. V~4MK Qhaersat9ZBE3 Vora used apprs~w~~~ a%oct@ a8 virious uaistma Eoroko as assess hot~& driad. OC WQQ QVBSd Chat horaous gQ Sud 2' hours...KK Hmm 'tiaao ZXZ"d. XnCEICno ~. ?n~ cpa~ Vms::Hach. . UIadpicoKdSq. X6, ESP& S&~o~c. L947, Tha Eccaaacii"cLBCII @IE HacCai. 'La Ljy Uqrlaj. ':. J? QvaacaL EELaoabiaKagyp X. " HSR;. K7p EESCKQ~~UZp g. 8 X956. ParCELU HaIILaa Saciemi HHCaC THicaa...

  7. Wet/dry cooling tower and method

    DOE Patents [OSTI]

    Glicksman, Leon R. (Lynnfield, MA); Rohsenow, Warren R. (Waban, MA)

    1981-01-01T23:59:59.000Z

    A wet/dry cooling tower wherein a liquid to-be-cooled is flowed along channels of a corrugated open surface or the like, which surface is swept by cooling air. The amount of the surface covered by the liquid is kept small compared to the dry part thereof so that said dry part acts as a fin for the wet part for heat dissipation.

  8. Energy storage connection system

    DOE Patents [OSTI]

    Benedict, Eric L.; Borland, Nicholas P.; Dale, Magdelena; Freeman, Belvin; Kite, Kim A.; Petter, Jeffrey K.; Taylor, Brendan F.

    2012-07-03T23:59:59.000Z

    A power system for connecting a variable voltage power source, such as a power controller, with a plurality of energy storage devices, at least two of which have a different initial voltage than the output voltage of the variable voltage power source. The power system includes a controller that increases the output voltage of the variable voltage power source. When such output voltage is substantially equal to the initial voltage of a first one of the energy storage devices, the controller sends a signal that causes a switch to connect the variable voltage power source with the first one of the energy storage devices. The controller then causes the output voltage of the variable voltage power source to continue increasing. When the output voltage is substantially equal to the initial voltage of a second one of the energy storage devices, the controller sends a signal that causes a switch to connect the variable voltage power source with the second one of the energy storage devices.

  9. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Tyacke, M.

    1993-08-01T23:59:59.000Z

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  10. Using and Storing Nonfat Dry Milk Nonfat dry milk is convenient to store, easy to use and

    E-Print Network [OSTI]

    in a cool, dry place. s Dry milk products are very sensitive to temperature and humidity. The area where your dry milk is stored should be kept as cool as possible. s Dry milk will absorb moisture and odorsUsing and Storing Nonfat Dry Milk Nonfat dry milk is convenient to store, easy to use

  11. Thermal Modeling Studies for Active Storage Modules in the Calvert Cliffs ISFSI

    SciTech Connect (OSTI)

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

    2013-06-14T23:59:59.000Z

    Temperature measurements obtained for two storage modules in the Calvert Cliffs Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI) as part of the Used Fuel Disposition Campaign of the Department of Energy (DOE) were used to perform validation and sensitivity studies on detailed computational fluid dynamics (CFD) models of the concrete storage modules, including the dry storage canister within the modules. The storage modules in the Calvert Cliffs Nuclear Power Station’s ISFSI are a site-specific version of the standard NUHOMS® HSM. The two modules inspected each contained a 24P DSC loaded with 24 CE 14x14 spent fuel assemblies. The thermal analysis was performed using the STAR-CCM+ package, and the models developed for the specific ISFSI modules yielded temperature predictions in actual storage conditions for the concrete structure, the DSC and its contents, including preliminary estimates of fuel cladding temperatures for the used nuclear fuel. The results of this work demonstrate that existing CFD modeling tools can be used to obtain reasonable and accurate detailed representations of spent fuel storage systems with realistic decay heat loadings when the model omits specific conservatisms and bounding assumptions normally used in design-basis and safety-basis calculations. This paper presents sensitivity studies on modeling detail (for the storage module and the DSC), boundary conditions, and decay heat load, to evaluate the effect of the modeling approach on predicted temperatures and temperature distributions. Because nearly all degradation mechanisms for materials and structures comprising dry storage and transportation systems are dependent on temperature, accurate characterization of local temperatures and temperature gradients that the various components of these systems will experience over the entire storage period has been identified as a primary requirement for evaluation of very long term storage of used nuclear fuel.

  12. FINAL REPORT: Transformational electrode drying process

    SciTech Connect (OSTI)

    Claus Daniel, C.; Wixom, M. (A123 Systems, Inc.)

    2013-12-19T23:59:59.000Z

    This report includes major findings and outlook from the transformational electrode drying project performance period from January 6, 2012 to August 1, 2012. Electrode drying before cell assembly is an operational bottleneck in battery manufacturing due to long drying times and batch processing. Water taken up during shipment and other manufacturing steps needs to be removed before final battery assembly. Conventional vacuum ovens are limited in drying speed due to a temperature threshold needed to avoid damaging polymer components in the composite electrode. Roll to roll operation and alternative treatments can increase the water desorption and removal rate without overheating and damaging other components in the composite electrode, thus considerably reducing drying time and energy use. The objective of this project was the development of an electrode drying procedure, and the demonstration of processes with no decrease in battery performance. The benchmark for all drying data was an 80°C vacuum furnace treatment with a residence time of 18 – 22 hours. This report demonstrates an alternative roll to roll drying process with a 500-fold improvement in drying time down to 2 minutes and consumption of only 30% of the energy compared to vacuum furnace treatment.

  13. Cold vacuum drying facility design requirements

    SciTech Connect (OSTI)

    IRWIN, J.J.

    1999-07-01T23:59:59.000Z

    This document provides the detailed design requirements for the Spent Nuclear Fuel Project Cold Vacuum Drying Facility. Process, safety, and quality assurance requirements and interfaces are specified.

  14. CHEMICAL STORAGE: MYTHS VERSUS REALITY

    SciTech Connect (OSTI)

    Simmons, F

    2007-03-19T23:59:59.000Z

    A large number of resources explaining proper chemical storage are available. These resources include books, databases/tables, and articles that explain various aspects of chemical storage including compatible chemical storage, signage, and regulatory requirements. Another source is the chemical manufacturer or distributor who provides storage information in the form of icons or color coding schemes on container labels. Despite the availability of these resources, chemical accidents stemming from improper storage, according to recent reports (1) (2), make up almost 25% of all chemical accidents. This relatively high percentage of chemical storage accidents suggests that these publications and color coding schemes although helpful, still provide incomplete information that may not completely mitigate storage risks. This manuscript will explore some ways published storage information may be incomplete, examine the associated risks, and suggest methods to help further eliminate chemical storage risks.

  15. Cask weeping mitigation

    DOE Patents [OSTI]

    Krumhansl, James L. (Albuquerque, NM); Brady, Patrick V. (Albuquerque, NM); Teter, David M. (Edgewood, NM); McConnell, Paul (Albuquerque, NM)

    2007-09-18T23:59:59.000Z

    A method (and concomitant kit) for treating a surface to reduce subsequent .sup.137Cs nuclide desorption comprising contacting the surface with a first cation-containing solution, the cation being one or more of Cs.sup.+, Rb.sup.+, Ag.sup.+, Tl.sup.+, K.sup.+, and NH.sub.4.sup.+, and contacting the surface with a second cation-containing solution, the cation being one or more of Cs.sup.+, Rb.sup.+, Ag.sup.+, Tl.sup.+, K.sup.+, and NH.sub.4.sup.+, thereby reducing amounts of radioactive cesium embedded in clays found on the surface.

  16. Storage Ring Parameters

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del Sol HomeFacebookScholarshipSpiralingSecurity217,354 217,814 218,494StorageStorage

  17. Precipitation scavenging, dry deposition, and resuspension. Volume 2: dry deposition and resuspension

    SciTech Connect (OSTI)

    Pruppacher, H.R.; Semanin, R.G.; Slinn, W.G.N.

    1983-01-01T23:59:59.000Z

    Papers are presented under the headings: dry deposition of gases, dry deposition of particles, wind erosion, plutonium deposition and resuspension, air-sea exchange, tropical and polar, global scale, and future studies.

  18. Sandia National Laboratories: Energy Storage Systems

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sandian Spoke at the New York Energy Storage Expo On December 12, 2014, in Energy, Energy Storage, Energy Storage Systems, Grid Integration, Infrastructure Security, News, News &...

  19. AQUIFER THERMAL ENERGY STORAGE-A SURVEY

    E-Print Network [OSTI]

    Tsang, Chin Fu

    2012-01-01T23:59:59.000Z

    High temperature underground thermal energy storage, inProceedings, Thermal Energy Storage in Aquifers Workshop:underground thermal energy storage, in ATES newsletter:

  20. THERMAL ENERGY STORAGE IN AQUIFERS WORKSHOP

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    Survey of Thermal Energy Storage in Aquifers Coupled withconcept of thermal energy storage in aquifers was suggestedLow Temperature Thermal Energy Storage Program of Oak Ridge

  1. Sandia National Laboratories: DOE International Energy Storage...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    International Energy Storage Database Has Logged 420 Energy Storage Projects Worldwide with 123 GW of Installed Capacity DOE International Energy Storage Database Has Logged 420...

  2. Distributed Generation with Heat Recovery and Storage

    E-Print Network [OSTI]

    Siddiqui, Afzal S.; Marnay, Chris; Firestone, Ryan M.; Zhou, Nan

    2008-01-01T23:59:59.000Z

    tiles for thermal energy storage,” working paper, Colorado1991). Wallboard with latent heat storage for passive solarR. (2000). Thermal energy storage for space cooling, Pacific

  3. Sandia National Laboratories: Electricity Storage Handbook

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Electricity Storage Handbook Published On July 31, 2013, in Energy, Energy Assurance, Energy Storage, Energy Storage Systems, Energy Surety, Grid Integration, Infrastructure...

  4. Nanostructured Materials for Energy Generation and Storage

    E-Print Network [OSTI]

    Khan, Javed Miller

    2012-01-01T23:59:59.000Z

    for Electrochemical Energy Storage Nanostructured ElectrodesCells for Energy Storage and Generation . . . . . . . . . .batteries and their energy storage efficiency. vii Contents

  5. NERSC Frontiers in Advanced Storage Technology Project

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Storage R&D Frontiers in Advanced Storage Technologies (FAST) project Working with vendors to develop new functionality in storage technologies generally not yet available to...

  6. THERMAL ENERGY STORAGE IN AQUIFERS WORKSHOP

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    Survey of Thermal Energy Storage in Aquifers Coupled withAnnual Thermal Energy Storage Contractors' InformationLarge-Scale Thermal Energy Storage for Cogeneration and

  7. Nanostructured Materials for Energy Generation and Storage

    E-Print Network [OSTI]

    Khan, Javed Miller

    2012-01-01T23:59:59.000Z

    of new energy generation and storage technologies arenew energy generation and storage technologies is importantBased Energy Storage and Generation Technologies The world

  8. Water Heaters (Storage Oil) | Department of Energy

    Energy Savers [EERE]

    Oil) Water Heaters (Storage Oil) Water Heater, Storage Oil - v1.0.xlsx More Documents & Publications Water Heaters (Tankless Electric) Water Heaters (Storage Electric)...

  9. AQUIFER THERMAL ENERGY STORAGE-A SURVEY

    E-Print Network [OSTI]

    Tsang, Chin Fu

    2012-01-01T23:59:59.000Z

    1978, High temperature underground thermal energy storage,in Proceedings, Thermal Energy Storage in Aquifers Workshop:High temperature underground thermal energy storage, in ATES

  10. NV Energy Electricity Storage Valuation

    SciTech Connect (OSTI)

    Ellison, James F.; Bhatnagar, Dhruv; Samaan, Nader A.; Jin, Chunlian

    2013-06-30T23:59:59.000Z

    This study examines how grid-level electricity storage may benet the operations of NV Energy in 2020, and assesses whether those benets justify the cost of the storage system. In order to determine how grid-level storage might impact NV Energy, an hourly production cost model of the Nevada Balancing Authority (\\BA") as projected for 2020 was built and used for the study. Storage facilities were found to add value primarily by providing reserve. Value provided by the provision of time-of-day shifting was found to be limited. If regulating reserve from storage is valued the same as that from slower ramp rate resources, then it appears that a reciprocating engine generator could provide additional capacity at a lower cost than a pumped storage hydro plant or large storage capacity battery system. In addition, a 25-MW battery storage facility would need to cost $650/kW or less in order to produce a positive Net Present Value (NPV). However, if regulating reserve provided by storage is considered to be more useful to the grid than that from slower ramp rate resources, then a grid-level storage facility may have a positive NPV even at today's storage system capital costs. The value of having storage provide services beyond reserve and time-of-day shifting was not assessed in this study, and was therefore not included in storage cost-benefit calculations.

  11. Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.

  12. Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495

    SciTech Connect (OSTI)

    Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine)] [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine); Schmieman, Eric [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)] [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)

    2013-07-01T23:59:59.000Z

    Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

  13. Effects of Sample Storage on Biosolids Compost Stability and Maturity Evaluation

    E-Print Network [OSTI]

    Ma, Lena

    Effects of Sample Storage on Biosolids Compost Stability and Maturity Evaluation L. Wu and L. Q. Ma-drying of soil sam-Compost stability and maturity are important parameters of com- ples is the most common practice (Bates, 1993), despitepost quality. To date, nearly all compost characterization has been

  14. The development of a management strategy for interim storage and final disposal of nuclear wastes

    SciTech Connect (OSTI)

    Engelmann, H.J.; Popp, F.W. [Deutsche Gesellschaft zum Bau and Betrieb von Endglagern fuer Abfallostofe mbH, Peine (Germany); Arntzen, P.; Botzem, W. [NUKEM GmbH, Alzenau (Germany); Soucek, B. [Czech Power Board, Prague (Czech Republic)

    1993-12-31T23:59:59.000Z

    The overall waste management strategy for alternative routes from reactor to final disposal, including dry interim storage, is discussed. Within the framework of a preliminary structure plan possible technical solutions must be investigated, and with sufficient relevant information available the future progress of the project, can be addressed on the base of a decision analysis.

  15. An experimental investigation of high temperature, high pressure paper drying

    E-Print Network [OSTI]

    Patel, Kamal Raoji

    1994-01-01T23:59:59.000Z

    % moisture removed oven dried mass of handsheet, g mass of handsheet after drying test, g mass of handsheet before drying test, g relative moisture removed from handsheet moisture removed by drying, % initial moisture (im) initial handsheet sample mass..., and the effects on the paper sheet and drying felt can be detrimental. Elevated temperatures reduce water viscosity which permits reduced resistance to water flow in the sheet. Pressing with a drying temperature of 95 C gives increased drying capacity, reduced...

  16. Infrared Dry-peeling Technology for Tomatoes

    E-Print Network [OSTI]

    Infrared Dry-peeling Technology for Tomatoes Saves Energy Energy Efficiency Research Office PIER This research will use infrared heating technology for peeling tomatoes. Infrared dry peeling, a device, producing less wastewater and preserving product quality. Infrared drypeeling is expected to reduce

  17. Cooking and Using Dried Beans and Peas

    E-Print Network [OSTI]

    Cooking and Using Dried Beans and Peas Beans and peas are good for you Beans and peas beans with rice or corn to provide high quality complete protein. If you are on a special diet, remember that beans and peas are low in sodium and fat. How to store dried beans and peas Store beans and peas

  18. Growing Dry Beans for an Emerging Market

    E-Print Network [OSTI]

    Hayden, Nancy J.

    Growing Dry Beans for an Emerging Market JOIN US FOR AN EVENING WITH JACK LAZOR, OF BUTTERWORKS FARM AND JOE BOSSEN, OF VERMONT BEAN CRAFTERS APRIL 10TH , 2012, 6:15-8PM AT THE KELLOGG-HUBBARD LIBRARY EAST MONTPELIER ROOM 135 MAIN ST., MONTPELIER, VT 05602 Jack Lazor has grown dry beans for local

  19. Underground pumped hydroelectric storage

    SciTech Connect (OSTI)

    Allen, R.D.; Doherty, T.J.; Kannberg, L.D.

    1984-07-01T23:59:59.000Z

    Underground pumped hydroelectric energy storage was conceived as a modification of surface pumped storage to eliminate dependence upon fortuitous topography, provide higher hydraulic heads, and reduce environmental concerns. A UPHS plant offers substantial savings in investment cost over coal-fired cycling plants and savings in system production costs over gas turbines. Potential location near load centers lowers transmission costs and line losses. Environmental impact is less than that for a coal-fired cycling plant. The inherent benefits include those of all pumped storage (i.e., rapid load response, emergency capacity, improvement in efficiency as pumps improve, and capacity for voltage regulation). A UPHS plant would be powered by either a coal-fired or nuclear baseload plant. The economic capacity of a UPHS plant would be in the range of 1000 to 3000 MW. This storage level is compatible with the load-leveling requirements of a greater metropolitan area with population of 1 million or more. The technical feasibility of UPHS depends upon excavation of a subterranean powerhouse cavern and reservoir caverns within a competent, impervious rock formation, and upon selection of reliable and efficient turbomachinery - pump-turbines and motor-generators - all remotely operable.

  20. Hydrogen Storage CODES & STANDARDS

    E-Print Network [OSTI]

    automotive start-up. · Air/Thermal/Water Management ­ improved air systems, high temperature membranes, heat to pump Hydrogen Fuel/ Storage/ Infrastructure $45/kW (2010) $30kW (2015) 325 W/kg 220 W/L 60% (hydrogen system Component Air management, sensors, MEA's, membranes, Bipolar Plates, fuel processor reactor zones

  1. Storage Ring | Advanced Photon Source

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    The Electron Storage Ring The 7-GeV electrons are injected into the 1104-m-circumference storage ring, a circle of more than 1,000 electromagnets and associated equipment, located...

  2. Chit-based Remote Storage

    E-Print Network [OSTI]

    Paluska, Justin Mazzola

    We propose a model for reliable remote storage founded on contract law. Consumers submit their bits to storage providers in exchange for a chit. A chit is a cryptographically secure, verifiable contract between a consumer ...

  3. Gaseous and Liquid Hydrogen Storage

    Broader source: Energy.gov [DOE]

    Today's state of the art for hydrogen storage includes 5,000- and 10,000-psi compressed gas tanks and cryogenic liquid hydrogen tanks for on-board hydrogen storage.

  4. Drying results of K-Basin fuel element 0309M (Run 3)

    SciTech Connect (OSTI)

    Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01T23:59:59.000Z

    An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step.

  5. Webinar: Hydrogen Storage Materials Requirements

    Broader source: Energy.gov [DOE]

    Video recording and text version of the webinar titled, Hydrogen Storage Materials Requirements, originally presented on June 25, 2013.

  6. The Power of Energy Storage

    E-Print Network [OSTI]

    Sadoulet, Elisabeth

    The Power of Energy Storage How to Increase Deployment in California to Reduce Greenhouse Gas;1Berkeley Law \\ UCLA Law The Power of Energy Storage: How to Increase Deployment in California to Reduce Greenhouse Gas Emissions Executive Summary: Expanding Energy Storage in California Sunshine and wind, even

  7. HIERARCHICAL STORAGE SYSTEMS FOR INTERACTIVE

    E-Print Network [OSTI]

    Tobagi, Fouad

    HIERARCHICAL STORAGE SYSTEMS FOR INTERACTIVE VIDEO­ON­DEMAND Shueng­Han Gary Chan and Fouad A; Hierarchical Storage Systems for Interactive Video­On­Demand Shueng­Han Gary Chan and Fouad A. Tobagi Technical­9040 pubs@shasta.stanford.edu Abstract On­demand video servers based on hierarchical storage systems

  8. GETTING CARBON CAPTURE AND STORAGE

    E-Print Network [OSTI]

    Haszeldine, Stuart

    GETTING CARBON CAPTURE AND STORAGE TECHNOLOGIES TO MARKET BREAKING THE DEADLOCK Report of a Science: Carbon Capture and Storage © OECD/IEA 2009, fig. 1, p. 6 Figures 2 and 3 reprinted with permission from `UK Carbon storage and capture, where is it?' by Stuart Haszeldine, Professor of Carbon Capture

  9. Normal matter storage of antiprotons

    SciTech Connect (OSTI)

    Campbell, L.J.

    1987-01-01T23:59:59.000Z

    Various simple issues connected with the possible storage of anti p in relative proximity to normal matter are discussed. Although equilibrium storage looks to be impossible, condensed matter systems are sufficiently rich and controllable that nonequilibrium storage is well worth pursuing. Experiments to elucidate the anti p interactions with normal matter are suggested. 32 refs.

  10. Above Ground Storage Tank (AST) Inspection Form

    E-Print Network [OSTI]

    Pawlowski, Wojtek

    Above Ground Storage Tank (AST) Inspection Form Petroleum Bulk Storage Form Facility Name.ehs.cornell.edu/env/bulk-material-storage/petroleum-bulk-storage/Documents/AST_Inspection_Form.pdf #12;

  11. Panel 4, Hydrogen Energy Storage Policy Considerations

    Broader source: Energy.gov (indexed) [DOE]

    Energy Storage Policy Considerations Hydrogen Storage Workshop Jeffrey Reed Southern California Gas Company May 15, 2014 0 Methane is a Great Storage Medium 1 SoCalGas' storage...

  12. 1.2.1.1 Harvest, Collection and Storage Quarter 3 Milestone Report

    SciTech Connect (OSTI)

    Lynn M Wendt; William A Smith; Kara G Cafferty; Ian J Bonner; Qiyang Huang; Rachel D Colby

    2014-07-01T23:59:59.000Z

    Single pass baling of corn stover is required in order to meet targets for the herbaceous biomass 2017 logistics design case. Single-pass pass stover harvest is based on the grain harvest and generally results in stover with a moisture content of 30-50% wet basis (w.b). Aerobic storage of corn stover with high moisture results in high levels of dry matter loss (DML), up to 25%. Anaerobic storage (ensiling) reduces DML to less than 5%, but additional costs are associated with handling and transporting the extra moisture in the biomass. This milestone provides a best-estimate of costs for using high moisture feedstock within the conventional baled logistics system. The costs of three (3) anaerobic storage systems that reduce dry matter losses (bale wrap, silage tube, and silage drive over pile) are detailed in this milestone and compared to both a conventional dry-baled corn stover case and a high moisture bale case, both stored aerobically. The total logistics cost (harvest, collection, storage, and transportation) of the scenarios are as follows: the conventional multi-pass dry bale case and the single-pass high moisture case stored aerobically were nearly equivalent at $61.15 and $61.24/DMT. The single-pass bale wrap case was the lowest at $57.63/DMT. The bulk anaerobic cases were the most expensive at $84.33 for the silage tube case and $75.97 for the drive over pile, which reflect the additional expense of transporting high-moisture bulk material; however, a reduction in preprocessing costs may occur because these feedstocks are size reduced in the field. In summary, the costs estimates presented in this milestone report can be used to determine if anaerobic storage of high-moisture corn stover is an economical option for dry matter preservation.

  13. Hot Dry Rock; Geothermal Energy

    SciTech Connect (OSTI)

    None

    1990-01-01T23:59:59.000Z

    The commercial utilization of geothermal energy forms the basis of the largest renewable energy industry in the world. More than 5000 Mw of electrical power are currently in production from approximately 210 plants and 10 000 Mw thermal are used in direct use processes. The majority of these systems are located in the well defined geothermal generally associated with crustal plate boundaries or hot spots. The essential requirements of high subsurface temperature with huge volumes of exploitable fluids, coupled to environmental and market factors, limit the choice of suitable sites significantly. The Hot Dry Rock (HDR) concept at any depth originally offered a dream of unlimited expansion for the geothermal industry by relaxing the location constraints by drilling deep enough to reach adequate temperatures. Now, after 20 years intensive work by international teams and expenditures of more than $250 million, it is vital to review the position of HDR in relation to the established geothermal industry. The HDR resource is merely a body of rock at elevated temperatures with insufficient fluids in place to enable the heat to be extracted without the need for injection wells. All of the major field experiments in HDR have shown that the natural fracture systems form the heat transfer surfaces and that it is these fractures that must be for geothermal systems producing from naturally fractured formations provide a basis for directing the forthcoming but, equally, they require accepting significant location constraints on HDR for the time being. This paper presents a model HDR system designed for commercial operations in the UK and uses production data from hydrothermal systems in Japan and the USA to demonstrate the reservoir performance requirements for viable operations. It is shown that these characteristics are not likely to be achieved in host rocks without stimulation processes. However, the long term goal of artificial geothermal systems developed by systematic engineering procedures at depth may still be attained if high temperature sites with extensive fracturing are developed or exploited. [DJE -2005

  14. Sandia National Laboratories: Energy Storage Systems

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Reserve University On January 28, 2014, in Computational Modeling & Simulation, Energy, Energy Storage, Energy Storage Systems, Infrastructure Security, Materials Science,...

  15. Sandia National Laboratories: Energy Storage Systems

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    in Center for Infrastructure Research and Innovation (CIRI), Energy, Energy Assurance, Energy Storage, Energy Storage Systems, Facilities, Infrastructure Security, Materials...

  16. Project Profile: Thermochemical Storage with Anhydrous Ammonia...

    Office of Environmental Management (EM)

    Storage with Anhydrous Ammonia: Optimizing the Synthesis Reactor for Direct Production of Supercritical Steam Project Profile: Thermochemical Storage with Anhydrous...

  17. THERMAL ENERGY STORAGE IN AQUIFERS WORKSHOP

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    1975. Underground Storage of Treated Water: A Field Test.1975. "Underground Storage of Treated Water: A Field Test,"

  18. Hydrogen Compression, Storage, and Dispensing Cost Reduction...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Compression, Storage, and Dispensing Cost Reduction Workshop Addendum Hydrogen Compression, Storage, and Dispensing Cost Reduction Workshop Addendum Document states additional...

  19. Combinatorial Approaches for Hydrogen Storage Materials (presentation...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Approaches for Hydrogen Storage Materials (presentation) Combinatorial Approaches for Hydrogen Storage Materials (presentation) Presentation on NIST Combinatorial Methods at the...

  20. Drying results of K-Basin fuel element 1164M (run 6)

    SciTech Connect (OSTI)

    Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-08-01T23:59:59.000Z

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the sixth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 1164 M. This element (referred to as Element 1164M) was stored underwater in the K-West Basin from 1983 until 1996. Element 1164M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0.

  1. Drying results of K-Basin fuel element 5744U (Run 4)

    SciTech Connect (OSTI)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01T23:59:59.000Z

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fourth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 5744U. This element (referred to as Element 5744U) was stored underwater in the K-West Basin from 1983 until 1996. Element 5744U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0.

  2. Acoustically enhanced heat exchange and drying apparatus

    DOE Patents [OSTI]

    Bramlette, T.T.; Keller, J.O.

    1987-07-10T23:59:59.000Z

    A heat transfer drying apparatus includes an acoustically augmented heat transfer chamber for receiving material to be dried. The chamber includes a first heat transfer gas inlet, a second heat transfer gas inlet, a material inlet, and a gas outlet which also serves as a dried material and gas outlet. A non-pulsing first heat transfer gas source provides a first drying gas to the acoustically augmented heat transfer chamber through the first heat transfer gas inlet. A valveless, continuous second heat transfer gas source provides a second drying gas to the acoustically augmented heat transfer chamber through the second heat transfer gas inlet. The second drying gas also generates acoustic waves which bring about acoustical coupling with the gases in the acoustically augmented heat transfer chamber. The second drying gas itself oscillates at an acoustic frequency of approximately 180 Hz due to fluid mechanical motion in the gas. The oscillations of the second heat transfer gas coupled to the first heat transfer gas in the acoustically augmented heat transfer chamber enhance heat and mass transfer by convection within the chamber. 3 figs.

  3. Thermoelectric powered wireless sensors for spent fuel monitoring

    SciTech Connect (OSTI)

    Carstens, T.; Corradini, M.; Blanchard, J. [Dept. of Engineering Physics, Univ. of Wisconsin-Madison, Madison, WI 53706 (United States); Ma, Z. [Dept. of Electrical and Computer Engineering, Univ. of Wisconsin-Madison, Madison, WI 53706 (United States)

    2011-07-01T23:59:59.000Z

    This paper describes using thermoelectric generators to power wireless sensors to monitor spent nuclear fuel during dry-cask storage. OrigenArp was used to determine the decay heat of the spent fuel at different times during the service life of the dry-cask. The Engineering Equation Solver computer program modeled the temperatures inside the spent fuel storage facility during its service life. The temperature distribution in a thermoelectric generator and heat sink was calculated using the computer program Finite Element Heat Transfer. From these temperature distributions the power produced by the thermoelectric generator was determined as a function of the service life of the dry-cask. In addition, an estimation of the path loss experienced by the wireless signal can be made based on materials and thickness of the structure. Once the path loss is known, the transmission power and thermoelectric generator power requirements can be determined. This analysis estimates that a thermoelectric generator can produce enough power for a sensor to function and transmit data from inside the dry-cask throughout its service life. (authors)

  4. Nanoscale data storage

    E-Print Network [OSTI]

    J. C. Li

    2007-01-29T23:59:59.000Z

    The object of this article is to review the development of ultrahigh-density, nanoscale data storage, i.e., nanostorage. As a fundamentally new type of storage system, the recording mechanisms of nanostorage may be completely different to those of the traditional devices. Currently, two types of molecules are being studied for potential application in nanostorage. One is molecular electronic elements including molecular wires, rectifiers, switches, and transistors. The other approach employs nanostructured materials such as nanotubes, nanowires, and nanoparticles. The challenges for nanostorage are not only the materials, ultrahigh data-densities, fabrication-costs, device operating temperatures and large-scale integration, but also the development of the physical principles and models. There are already some breakthroughs obtained, but it is still unclear what kind of nanostorage systems can ultimately replace the current silicon based transistors. A promising candidate may be a molecular-nanostructure hybrid device with sub-5 nm dimensions.

  5. Superconducting magnetic energy storage

    SciTech Connect (OSTI)

    Hassenzahl, W.

    1988-08-01T23:59:59.000Z

    Recent programmatic developments in Superconducting Magnetic Energy Storage (SMES) have prompted renewed and widespread interest in this field. In mid 1987 the Defense Nuclear Agency, acting for the Strategic Defense Initiative Office, issued a request for proposals for the design and construction of SMES Engineering Test Model (ETM). Two teams, one led by Bechtel and the other by Ebasco, are now engaged in the first phase of the development of a 10 to 20 MWhr ETM. This report presents the rationale for energy storage on utility systems, describes the general technology of SMES, and explains the chronological development of the technology. The present ETM program is outlined; details of the two projects for ETM development are described in other papers in these proceedings. The impact of high T/sub c/ materials on SMES is discussed. 69 refs., 3 figs., 3 tabs.

  6. HYDROGEN USAGE AND STORAGE

    E-Print Network [OSTI]

    It is thought that it will be useful to inform society and people who are interested in hydrogen energy. The study below has been prepared due to this aim can be accepted as an article to exchange of information between people working on this subject. This study has been presented to reader to be utilized as a “technical note”. Main Energy sources coal, petroleum and natural gas are the fossil fuels we use today. They are going to be exhausted since careless usage in last decades through out the world, and human being is going to face the lack of energy sources in the near future. On the other hand as the fossil fuels pollute the environment makes the hydrogen important for an alternative energy source against to the fossil fuels. Due to the slow progress in hydrogen’s production, storage and converting into electrical energy experience, extensive usage of Hydrogen can not find chance for applications in wide technological practices. Hydrogen storage stands on an important point in the development of Hydrogen energy Technologies. Hydrogen is volumetrically low energy concentration fuel. Hydrogen energy, to meet the energy quantity necessary for the nowadays technologies and to be accepted economically and physically against fossil fuels, Hydrogen storage technologies have to be developed in this manner. Today the most common method in hydrogen storage may be accepted as the high pressurized composite tanks. Hydrogen is stored as liquid or gaseous phases. Liquid hydrogen phase can be stored by using composite tanks under very high pressure conditions. High technology composite material products which are durable to high pressures, which should not be affected by hydrogen embrittlement and chemical conditions.[1

  7. Maui energy storage study.

    SciTech Connect (OSTI)

    Ellison, James; Bhatnagar, Dhruv; Karlson, Benjamin

    2012-12-01T23:59:59.000Z

    This report investigates strategies to mitigate anticipated wind energy curtailment on Maui, with a focus on grid-level energy storage technology. The study team developed an hourly production cost model of the Maui Electric Company (MECO) system, with an expected 72 MW of wind generation and 15 MW of distributed photovoltaic (PV) generation in 2015, and used this model to investigate strategies that mitigate wind energy curtailment. It was found that storage projects can reduce both wind curtailment and the annual cost of producing power, and can do so in a cost-effective manner. Most of the savings achieved in these scenarios are not from replacing constant-cost diesel-fired generation with wind generation. Instead, the savings are achieved by the more efficient operation of the conventional units of the system. Using additional storage for spinning reserve enables the system to decrease the amount of spinning reserve provided by single-cycle units. This decreases the amount of generation from these units, which are often operated at their least efficient point (at minimum load). At the same time, the amount of spinning reserve from the efficient combined-cycle units also decreases, allowing these units to operate at higher, more efficient levels.

  8. Advanced wet-dry cooling tower concept

    E-Print Network [OSTI]

    Snyder, Troxell Kimmel

    The purpose of this years' work has been to test and analyze the new dry cooling tower surface previously developed. The model heat transfer test apparatus built last year has been instrumented for temperature, humidity ...

  9. Cold vacuum drying facility design requirements

    SciTech Connect (OSTI)

    Irwin, J.J.

    1997-09-24T23:59:59.000Z

    This release of the Design Requirements Document is a complete restructuring and rewrite to the document previously prepared and released for project W-441 to record the design basis for the design of the Cold Vacuum Drying Facility.

  10. Dry cooling: Perspectives on future needs

    SciTech Connect (OSTI)

    Guyer, E.C. (Yankee Scientific, Inc., Ashland, MA (United States))

    1991-08-01T23:59:59.000Z

    The factors that can be expected to determine the future role of dry cooling in the United States electric power generation industry are identified and characterized. Focus is primarily on the issues of water availability for the electric power industry and the environmental impacts of evaporative cooling systems. The question of future water availability is addressed in terms of both limitations and opportunities facing the industry. A brief review of the status of dry cooling applications is provided. Included is a summary of an extensive survey of electric utility industry perspectives on the future requirements and role for dry cooling. Some regional assessments of the expected future requirements for this technology are also provided. Conclusions are a qualitative characterization of the expected future role of dry cooling in the electric power industry. 72 refs., 7 figs., 13 tabs.

  11. Resuspension and dry deposition research needs

    SciTech Connect (OSTI)

    Sehmel, G.A.

    1983-01-01T23:59:59.000Z

    The author concludes that better predictive models are needed for the signifcant health, ecological, and economic impacts of resuspended particles and their subsequent dry deposition. Both chemical and radioactive aerosols are discussed. (PSB)

  12. High strength air-dried aerogels

    DOE Patents [OSTI]

    Coronado, Paul R.; Satcher, Jr., Joe H.

    2012-11-06T23:59:59.000Z

    A method for the preparation of high strength air-dried organic aerogels. The method involves the sol-gel polymerization of organic gel precursors, such as resorcinol with formaldehyde (RF) in aqueous solvents with R/C ratios greater than about 1000 and R/F ratios less than about 1:2.1. Using a procedure analogous to the preparation of resorcinol-formaldehyde (RF) aerogels, this approach generates wet gels that can be air dried at ambient temperatures and pressures. The method significantly reduces the time and/or energy required to produce a dried aerogel compared to conventional methods using either supercritical solvent extraction. The air dried gel exhibits typically less than 5% shrinkage.

  13. Groundwater and Terrestrial Water Storage

    E-Print Network [OSTI]

    Rodell, M; Chambers, D P; Famiglietti, J S

    2011-01-01T23:59:59.000Z

    Lebanon, Syria, western Kazakhstan, Armenia, Georgia, andthe south Caucasus and west Kazakhstan were dry. Armenia hadwere detected over Russia and Kazakhstan. The anomaly lasted

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

    SciTech Connect (OSTI)

    Easton, Earl; Bajwa, Christopher; Li, Zhian; Gordon, Matthew [U.S. Nuclear Regulatory Commission, Washington, DC 20005 (United States)

    2012-07-01T23:59:59.000Z

    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)

  15. Compression of cooked freeze-dried carrots

    E-Print Network [OSTI]

    Macphearson, Bruce Alan

    1973-01-01T23:59:59.000Z

    to precompression characteristics (Brockmann, 1966). Hsmdy (1962) found that acceptable, compressed and freeze-dried spinach could be obtained by plasticizing the product to a moisture content of 9X before compression. Ishler (1962) reported that spraying... the dehydrated food before compression with either water, glycerine or propylene glycol produced bars with excellent rehydra- tion characteristics. He recommended spraying freeze-dried cellu- lar foods to 5-13X moisture, compressing, and redrying to lees than...

  16. Determination of Water Saturation in Relatively Dry Porous Media...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Water Saturation in Relatively Dry Porous Media Using Gas-phase Tracer Tests. Determination of Water Saturation in Relatively Dry Porous Media Using Gas-phase Tracer Tests....

  17. Wet-dry cooling demonstration. Test results

    SciTech Connect (OSTI)

    Allemann, R.T.; DeBellis, D.E.; Werry, E.V.; Johnson, B.M.

    1986-05-01T23:59:59.000Z

    A large-scale test of dry/wet cooling using the ammonia phase-change system, designated the Advanced Concepts Test (ACT), has been operated at Pacific Gas and Electric Company's Kern Station at Bakersfield, California. The facility is capable of condensing 60,000 lbs/h of steam from a small house turbine. Two different modes of combining dry and evaporative cooling have been tested. One uses deluge cooling in which water is allowed to flow over the fins of the dry (air-cooled) heat exchanger on hot days; the other uses a separate evaporative condenser in parallel to the dry heat exchanger. A third mode of enhancing the dry cooling system, termed capacitive cooling has been tested. In this system, the ammonia-cooled steam condenser is supplemented by a parallel conventional water-cooled condenser with water supplied from a closed system. This water is cooled during off-peak hours each night by an ammonia heat pump which rejects heat through the ACT Cooling Tower. If operated over the period of a year, each of the wet/dry systems would use only 25% of the water normally required to reject this heat load in an evaporative cooling tower. The third would consume no water, the evaporative cooling being replaced by the delayed cooling of the closed system water supply.

  18. Steam atmosphere drying exhaust steam recompression system

    DOE Patents [OSTI]

    Becker, F.E.; Smolensky, L.A.; Doyle, E.F.; DiBella, F.A.

    1994-03-08T23:59:59.000Z

    This invention relates to a heated steam atmosphere drying system comprising dryer in combination with an exhaust recompression system which is extremely energy efficient and eliminates dangers known to air dryers. The system uses superheated steam as the drying medium, which recirculates through the system where its heat of evaporation and heat of compression is recovered, thereby providing a constant source of heat to the drying chamber. The dryer has inlets whereby feedstock and superheated steam are fed therein. High heat transfer and drying rates are achieved by intimate contact of the superheated steam with the particles being dried. The dryer comprises a vessel which enables the feedstock and steam to enter and recirculate together. When the feedstock becomes dry it will exit the dryer with the steam and become separated from the steam through the use of a curvilinear louver separator (CLS). The CLS enables removal of fine and ultrafine particles from the dryer. Water vapor separated from the particles in the CLS as superheated steam, may then be recovered and recirculated as steam through the use of a compressor to either directly or indirectly heat the dryer, and a heat exchanger or a heater to directly provide heat to the dryer. This system not only provides a very efficient heat transfer system but results in a minimum carry-over of ultrafine particles thereby eliminating any explosive hazard. 17 figures.

  19. Steam atmosphere drying exhaust steam recompression system

    DOE Patents [OSTI]

    Becker, Frederick E. (Reading, MA); Smolensky, Leo A. (Concord, MA); Doyle, Edward F. (Dedham, MA); DiBella, Francis A. (Roslindale, MA)

    1994-01-01T23:59:59.000Z

    This invention relates to a heated steam atmosphere drying system comprising dryer in combination with an exhaust recompression system which is extremely energy efficient and eliminates dangers known to air dryers. The system uses superheated steam as the drying medium, which recirculated through the system where its heat of evaporation and heat of compression is recovered, thereby providing a constant source of heat to the drying chamber. The dryer has inlets whereby feedstock and superheated steam are fed therein. High heat transfer and drying rates are achieved by intimate contact of the superheated steam with the particles being dried The dryer comprises a vessel which enables the feedstock and steam to enter recirculate together. When the feedstock becomes dry it will exit the dryer with the steam and become separated from the steam through the use of a curvilinear louver separator (CLS). The CLS enables removal of fine and ultrafine particles from the dryer. Water vapor separated from the particles in the CLS as superheated steam, may then be recovered and recirculated as steam through the use of a compressor to either directly or indirectly heat the dryer, and a heat exchanger or a heater to directly provide heat to the dryer. This system not only provides a very efficient heat transfer system but results in a minimum carry-over of ultrafine particles thereby eliminating any explosive hazard.

  20. Rate of drying and stresses in the first period of drying

    SciTech Connect (OSTI)

    Kowalski, S.J.; Rybicki, A.

    2000-03-01T23:59:59.000Z

    The paper presents a computer simulated processes and illustrate how the drying induced stresses are influenced by the rate of drying. It is shown that the moisture transport coefficient, and thus the rate of drying, depends on the thermal state of the drying material, defined by the wet-bulb temperature. Through these simulated processes one can observe the evolution of the moisture content and stress distributions during drying at constant, but in each process different, wet-bulb temperatures. A convective drying process of a bar with rectangular cross-section is considered as example, and a two-dimensional initial-boundary value problem is solved numerically with the use of the finite element method. The numerical results are visualized in spatial diagrams.