Science.gov

Sample records for fuel dry interim

  1. Behavior of spent nuclear fuel and storage system components in dry interim storage. Revision 1

    SciTech Connect

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

    1983-02-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom; organic-cooled reactor (OCR) fuel (clad with a zirconium alloy) in silos in Canada; and boiling water reactor (BWR) fuel (clad with Zircaloy) in a metal storage cask in Germany. Dry storage demonstrations are under way for Zircaloy-clad fuel from BWRs, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions. 110 refs., 22 figs., 28 tabs.

  2. Behavior of spent nuclear fuel and storage system components in dry interim storage.

    SciTech Connect

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

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions.

  3. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  4. Interim Storage of Hanford Spent Fuel & Associated Sludge

    SciTech Connect

    MAKENAS, B.J.

    2002-07-01

    The Hanford site is currently dealing with a number of types of Spent Nuclear Fuel. The route to interim dry storage for the various fuel types branches along two different paths. Fuel types such as metallic N reactor fuel and Shippingport Core 2 Blanket assemblies are being placed in approximately 4 m long canisters which are then stored in tubes below grade in a new canister storage building. Other fuels such as TRIGA{trademark} and Light Water Reactor fuel will be relocated and stored in stand-alone casks on a concrete pad. Varying degrees of sophistication are being applied with respect to the drying and/or evacuation of the fuel interim storage canisters depending on the reactivity of the fuel, the degree of damaged fuel and the previous storage environment. The characterization of sludge from the Hanford K Basins is nearly complete and canisters are being designed to store the sludge (including uranium particles from fuel element cleaning) on an interim basis.

  5. Spent fuel behavior in dry storage

    NASA Astrophysics Data System (ADS)

    Johnson, A. B., Jr.; Pankaskie, P. J.; Gilbert, E. R.

    1982-02-01

    Dry storage is emerging as an attractive and timely alternative to complement wet storage, and assist utilities to meet interim storage needs. Spent fuel is handled and stored under dry conditions. Dry storage of irradiated Zircaloy clad fuel in metal casks, drywells, silos and vaults is demonstrated. Hot cell and laboratory studies also are underway to investigate specific phenomena related to cladding behavior in dry storage. A substantial fraction of the LWR spent fuel inventory has aged for relatively long times and has relatively low decay heats. This suggests that much of the fuel inventory can be stored at relatively low temperatures. Alternatively, rod consolidation of the older can be considered without exceeding maximum cladding temperatures.

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

    SciTech Connect

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

    1998-07-01

    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.

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

    SciTech Connect

    Lee, S.Y.

    1999-01-13

    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.

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

    SciTech Connect

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

    1998-07-01

    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

  9. Interim report spent nuclear fuel retrieval system fuel handling development testing

    SciTech Connect

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  10. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

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

    SciTech Connect

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M.

    2013-07-01

    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.

  12. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    SciTech Connect

    NORTON SH

    2010-02-23

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  13. INTERIM STORAGE AND LONG TERM DISPOSAL OF RESEARCH REACTOR SPENT FUEL

    SciTech Connect

    Vinson, D

    2006-08-22

    Aluminum clad research reactor spent nuclear fuel (SNF) is currently being consolidated in wet storage basins (pools). Approximately 20 metric tons (heavy metal) of aluminum-based spent nuclear fuel (Al-SNF) is being consolidated for treatment, packaging, interim storage, and preparation for ultimate disposal in a geologic repository. The storage and disposal of Al-SNF are subject to requirements that provide for safety and acceptable radionuclide release. The options studied for interim storage of SNF include wet storage and dry storage. Two options have also been studied to develop the technical basis for the qualification and repository disposal of aluminum spent fuel. The two options studied include Direct Disposal and Melt-Dilute treatment. The implementation of these options present relative benefits and challenges. Both the Direct Disposal and the Melt-Dilute treatment options have been developed and their technical viability assessed. Adaptation of the melt-dilute technology for the treatment of spent fuel offers the benefits of converting the spent fuel into a proliferation resistant form and/or significantly reducing the volume of the spent fuel. A Mobile Melt-Dilute system concept has emerged to realize these benefits and a prototype system developed. The application of the melt-dilute technology for the treatment of legacy nuclear materials has been evaluated and also offers the promise for the safe disposal of these materials.

  14. 78 FR 40199 - Draft Spent Fuel Storage and Transportation Interim Staff Guidance

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-03

    ... COMMISSION Draft Spent Fuel Storage and Transportation Interim Staff Guidance AGENCY: Nuclear Regulatory... Regulatory Commission (NRC) requests public comment on Draft Spent Fuel Storage and Transportation Interim... Integrity for Continued Storage of High Burnup Fuel Beyond 20 Years.'' The draft SFST-ISG provides...

  15. 75 FR 45678 - Notice of Availability of Interim Staff Guidance Document for Fuel Cycle Facilities

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-03

    ... COMMISSION Notice of Availability of Interim Staff Guidance Document for Fuel Cycle Facilities AGENCY... . SUPPLEMENTARY INFORMATION: I. Introduction The Nuclear Regulatory Commission (NRC) prepares and issues Interim... this notice is provided in the following table. ADAMS Interim staff guidance accession No. FCSS...

  16. Developing a concept for a national used fuel interim storage facility in the United States

    SciTech Connect

    Lewis, Donald Wayne

    2013-07-01

    In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a 'Monitored Retrievable Storage' facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to build a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE's goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility. (authors)

  17. Interim results from UO/sub 2/ fuel oxidation tests in air

    SciTech Connect

    Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j

    1987-08-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO/sub 2/, fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO/sub 2/ pellets in the temperature range of 135 to 250/sup 0/C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10/sup 5/ R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10/sup 5/ R/h gamma field. 33 refs., 51 figs., 6 tabs.

  18. Dry Transfer Systems for Used Nuclear Fuel

    SciTech Connect

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    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.

  19. Basis for assessing the movement of spent nuclear fuels from wet to dry storage at the Idaho Chemical Processing Plant

    SciTech Connect

    Guenther, R.J.; Gilbert, E.R.; Johnson, A.B.; Lund, A.L.; Pednekar, S.P.; Windes, W.E.

    1994-12-01

    An assessment of the possible material interactions arising from the movement of previously wet stored spent nuclear fuel (SNF) into long-term dry interim storage has been conducted for selected fuels in the Idaho Chemical Processing Plant (ICPP). Three main classes of fuels are addressed: aluminum (Al) clad, stainless steel (SS) clad, and unclad Uranium-Zirconium Hydride (UZrHx) fuel types. Degradation issues for the cladding, fuel matrix material, and storage canister in both wet and dry storage environments are assessed. Possible conditioning techniques to stabilize the fuel and optimum dry environment conditions during storage are also addressed.

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

    SciTech Connect

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

    2003-02-27

    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.

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

    SciTech Connect

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

    1982-09-01

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

  2. Spent Fuel Drying System Test Results (Dry-Run in Preparation for Run 8)

    SciTech Connect

    BM Oliver; GS Klinger; J Abrefah; SC Marschman; PJ MacFarlan; GA Ritter

    1999-08-11

    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 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL)(a)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of a test ''dry-run'' conducted prior to the eighth and last of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister6513U. The system used for the dry-run 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. The experimental results are provided in Section 4.0 and discussed Section 5.0.

  3. Dry Processing of Used Nuclear Fuel

    SciTech Connect

    K. M. Goff; M. F. Simpson

    2009-09-01

    Dry (non-aqueous) separations technologies have been used for treatment of used nuclear fuel since the 1960s, and they are still being developed and demonstrated in many countries. Dry technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. Within the Department of Energy’s Advanced Fuel Cycle Initiative, an electrochemical process employing molten salts is being developed for recycle of fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. Much of the development of this technology is based on treatment of used Experimental Breeder Reactor II (EBR-II) fuel, which is metallic. Electrochemical treatment of the EBR-II fuel has been ongoing in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory since 1996. More than 3.8 metric tons of heavy metal of metallic fast reactor fuel have been treated using this technology. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including high-level waste work. A historic perspective on the background of dry processing will also be provided.

  4. Fuel-Cell Structure Prevents Membrane Drying

    NASA Technical Reports Server (NTRS)

    Mcelroy, J.

    1986-01-01

    Embossed plates direct flows of reactants and coolant. Membrane-type fuel-cell battery has improved reactant flow and heat removal. Compact, lightweight battery produces high current and power without drying of membranes.

  5. Annotated Bibliography for Drying Nuclear Fuel

    SciTech Connect

    Rebecca E. Smith

    2011-09-01

    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.

  6. Inspection of Used Fuel Dry Storage Casks

    SciTech Connect

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

    2012-09-01

    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.

  7. Realization of the German Concept for Interim Storage of Spent Nuclear Fuel - Current Situation and Prospects

    SciTech Connect

    Thomauske, B. R.

    2003-02-25

    The German government has determined a phase out of nuclear power. With respect to the management of spent fuel it was decided to terminate transports to reprocessing plants by 2005 and to set up interim storage facilities on power plant sites. This paper gives an overview of the German concept for spent fuel management focused on the new on-site interim storage concept and the applied interim storage facilities. Since the end of the year 1998, the utilities have applied for permission of on-site interim storage in 13 storage facilities and 5 storage areas; one application for the interim storage facility Stade was withdrawn due to the planned final shut down of Stade nuclear power plant in autumn 2003. In 2001 and 2002, 3 on-site storage areas and 2 on-site storage facilities for spent fuel were licensed by the Federal Office for Radiation Protection (BfS). A main task in 2002 and 2003 has been the examination of the safety and security of the planned interim storage facilities and the verification of the licensing prerequisites. In the aftermath of September 11, 2001, BfS has also examined the attack with a big passenger airplane. Up to now, these aircraft crash analyses have been performed for three on-site interim storage facilities; the fundamental results will be presented. It is the objective of BfS to conclude the licensing procedures for the applied on-site interim storage facilities in 2003. With an assumed construction period for the storage buildings of about two years, the on-site interim storage facilities could then be available in the year 2005.

  8. Horizontal modular dry irradiated fuel storage system

    DOEpatents

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

    1988-01-01

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

  9. Interim Safety Basis for Fuel Supply Shutdown Facility

    SciTech Connect

    BENECKE, M.W.

    2000-09-07

    This ISB, in conjunction with the IOSR, provides the required basis for interim operation or restrictions on interim operations and administrative controls for the facility until a SAR is prepared in accordance with the new requirements or the facility is shut down. It is concluded that the risks associated with tha current and anticipated mode of the facility, uranium disposition, clean up, and transition activities required for permanent closure, are within risk guidelines.

  10. Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface.

    SciTech Connect

    Enos, David; Bryan, Charles R.

    2015-10-01

    Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

  11. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    SciTech Connect

    Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R.; Rossa, R.; Liljenfeldt, H.

    2015-07-01

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)

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

    SciTech Connect

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

    1996-09-01

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

  13. Fuel fire tests of selected assemblies. Interim report

    SciTech Connect

    Kydd, G.; Spindola, K.; Askew, G.K.

    1982-04-13

    A varing assortment of clothing assemblies was tested in the Fuel Fire Test Facility at the Naval Air Development Center. Included was a Nomex-Kevlar Cloque Coverall which had relatively good protection from fuel flames.

  14. Drying results of K-Basin fuel element 5744U (Run 4)

    SciTech Connect

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

    1998-07-01

    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.

  15. Drying Results of K-Basin Fuel Element 2660M (Run 7)

    SciTech Connect

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

    1999-07-26

    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 seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M 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.

  16. Drying Results of K-Basin Fuel Element 6513U (Run 8)

    SciTech Connect

    BM Oliver; GS Klinger; J Abrefah; SC Marschman; PJ MacFarlan; GA Ritter

    1999-08-11

    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 eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U 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.

  17. Drying results of K-Basin fuel element 1164M (run 6)

    SciTech Connect

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

    1998-08-01

    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.

  18. Drying Results of K-Basin Fuel Element 6603M (Rune 5)

    SciTech Connect

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

    1999-09-24

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels 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 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 fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which 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. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0.

  19. Drying results of K-Basin fuel element 1990 (Run 1)

    SciTech Connect

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

    1998-06-01

    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 fuels 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 first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element 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 testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.

  20. Identification of a breached fuel pin in the IEM (Interim Examination and Maintenance) cell

    SciTech Connect

    McGuinness, P.W.; Kalk, J.J.; Hicks, D.F.

    1987-01-01

    Novel methods were successfully employed to identify one breached fuel pin in a 217-pin fuel assembly. The assembly was an experiment that had been irradiated at the Fast Flux Test Facility (FFTF), an experimental liquid-metal reactor operated by Westinghouse Hanford Company for the US Department of Energy. A fuel assembly known to contain breached fuel pins was removed from the sodium-cooled FFTF reactor in November 1984. Later, this assembly was brought into the FFTF's Interim Examination and Maintenance (IEM) cell to be disassembled and, for the first time ever at FFTF, to identify a breached fuel pin. The synergistic evaluation of the four different verification techniques - visual examination, cladding swipe activity, wash water radiochemistry, and pin weight - provided rapid and positive identification. The capability to perform future detective work of this kind has been conclusively demonstrated.

  1. Drying studies of simulated DOE aluminum plate fuels

    SciTech Connect

    Lords, R.E.; Windes, W.E.; Crepeau, J.C.; Sidwell, R.W.

    1996-05-01

    Experiments have been conducted to validate the Idaho National Engineering Laboratory (INEL) drying procedures for preparation of corroded aluminum plate fuel for dry storage in an existing vented (and filtered) fuel storage facility. A mixture of hydrated aluminum oxide bound with a clay was used to model the aluminum corrosion product and sediment expected in these Department of Energy (DOE) owned fuel types. Previous studies demonstrated that the current drying procedures are adequate for removal of free water inside the storage canister and for transfer of this fuel to a vented dry storage facility. However, using these same drying procedures, the simulated corrosion product was found to be difficult to dry completely from between the aluminum clad plates of the fuel. Another related set of experiments was designed to ensure that the fuel would not be damaged during the drying process. Aluminum plate fuels are susceptible to pitting damage on the cladding that can result in a portion of UAl{sub x} fuel meat being disgorged. This would leave a water-filled void beneath the pit in the cladding. The question was whether bursting would occur when water in the void flashes to steam, causing separation of the cladding from the fuel, and/or possible rupture. Aluminum coupons were fabricated to model damaged fuel plates. These coupons do not rupture or sustain any visible damage during credible drying scenarios.

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

    SciTech Connect

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

    1999-12-01

    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

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

    SciTech Connect

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R

    1996-07-01

    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.

  4. Review of Drying Methods for Spent Nuclear Fuel

    SciTech Connect

    Large, W.S.

    1999-10-21

    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.

  5. Corrosion assessment of dry fuel storage containers

    SciTech Connect

    Graves, C.E.

    1994-09-01

    The structural stability as a function of expected corrosion degradation of 75 dry fuel storage containers located in the 200 Area Low-Level Waste Burial Grounds was evaluated. These containers include 22 concrete burial containers, 13 55-gal (208-l) drums, and 40 Experimental Breeder Reactor II (EBR-II) transport/storage casks. All containers are buried beneath at least 48 in. of soil and a heavy plastic tarp with the exception of 35 of the EBR-II casks which are exposed to atmosphere. A literature review revealed that little general corrosion is expected and pitting corrosion of the carbon steel used as the exterior shell for all containers (with the exception of the concrete containers) will occur at a maximum rate of 3.5 mil/yr. Penetration from pitting of the exterior shell of the 208-l drums and EBR-II casks is calculated to occur after 18 and 71 years of burial, respectively. The internal construction beneath the shell would be expected to preclude containment breach, however, for the drums and casks. The estimates for structural failure of the external shells, large-scale shell deterioration due to corrosion, are considerably longer, 39 and 150 years respectively for the drums and casks. The concrete burial containers are expected to withstand a service life of 50 years.

  6. Identification of a breached fuel pin in the Interim Examination and Maintenance Cell

    SciTech Connect

    McGuiness, P.W.; Kalk, J.J.; Hicks, D.F.

    1987-09-01

    At the Interim Examination and Maintenance (IEM) Cell, experiments are routinely disassembled and examined following irradiation in the Fast Flux Test Facility (FFTF). Recently and for the first time, a fueled experiment which had breached its cladding during irradiation was disassembled in the cell. The processing objective was to locate and identify the one pin (out of 217 pins) with breached cladding, and recover selected test pins for further examination. Identification of the breached pin proved to be challenging. After all pins were weighed the data were inconclusive, and alternate procedures had to be developed and implemented. Ultimately, four independent methods were used to pinpoint the breached pin.

  7. Three-dimensional Computational Fluid Dynamics (CFD) modeling of dry spent nuclear fuel storage canisters

    SciTech Connect

    Lee, S.Y.

    1997-06-01

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

  8. Preliminary Design Report Shippingport Spent Fuel Drying and Inerting System

    SciTech Connect

    JEPPSON, D.W.

    2000-05-18

    A process description and system flow sheets have been prepared to support the design/build package for the Shippingport Spent Fuel Canister drying and inerting process skid. A process flow diagram was prepared to show the general steps to dry and inert the Shippingport fuel loaded into SSFCs for transport and dry storage. Flow sheets have been prepared to show the flows and conditions for the various steps of the drying and inerting process. Calculations and data supporting the development of the flow sheets are included.

  9. Spent Fuel Test - Climax: technical measurements. Interim report, fiscal year 1982

    SciTech Connect

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

    1983-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted for the US Department of Energy (DOE) under the technical direction of the Lawrence Livermore National Laboratory (LLNL). Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April to May 1980, thus initiating a test with a planned 3- to 5-year fuel storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Three exchanges of spent fuel between the SFT-C and a surface storage facility furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and two previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 2-1/2 years of the test on more than 900 channels. Data continue to be acquired from the test. Some data are now available for analysis and are presented here. Highlights of activities this year include analysis of fracture data obtained during site characterization, laboratory studies of radiation effects and drilling damage in Climax granite, improved calculations of near-field heat transfer and thermomechanical response, a ventilation effects study, and further development of the data acquisition and management systems.

  10. Evaluation of gasification and gas cleanup processes for use in molten-carbonate fuel-cell power plants. Task B interim report

    SciTech Connect

    Not Available

    1981-12-01

    This interim report satisfies the Task B requirement for DOE Contract DE-AC21-81MC16220 to define process configurations for systems suitable for supplying fuel to molten carbonate fuel cells (MCFC) in industrial and utility power plants. The information and data necessary for this study were extracted from sources in the public domain, including reports from DOE, EPRI, and EPA; work sponsored in whole or in part by Federal agencies; and from trade journals, MCFC developers, and manufacturers. The configurations include entrained, fluidized-bed, gravitating-bed, and molten salt gasifiers, both air and oxygen blown. Desulfurization systems utilizing wet scrubbing processes, such as Selexol and Rectisol II, and dry sorbents, such as iron oxide and dolomite, were chosen for evaluation.

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

    SciTech Connect

    CHASTAIN, S.A.

    2005-10-24

    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

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

    SciTech Connect

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L.; Moore, E.N.

    2013-07-01

    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

  13. Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

    SciTech Connect

    Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Ganow, H.C.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

    1984-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables.

  14. Interim storage technology of spent fuel and high-level waste in Germany

    SciTech Connect

    Geiser, H.; Schroder, J.

    2007-07-01

    The idea of using casks for interim storage of spent fuel arose at GNS after a very controversial political discussion in 1978, when total passive safety features (including aircraft crash conditions) were required for an above ground spent fuel storage facility. In the meantime, GNS has loaded more than 1000 casks at 25 different storage sites in Germany. GNS cask technology is used in 13 countries. Spent fuel assemblies of PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high level waste containers are stored in full metal casks of the CASTOR{sup R} type. Also MOX fuel of PWR and BWR has been stored. More than two decades of storage have shown that the basic requirements (safe confinement, criticality safety, sufficient shielding and appropriate heat transfer) have been fulfilled in any case - during normal operation and in case of severe accidents, including aircraft crash. There is no indication of problems arising in the future. Of course, the experience of more than 20 years has resulted in improvements of the cask design. The CASTOR{sup R} casks have been thoroughly investigated by many experiments. There have been approx. 50 full and half scale drop tests and a significant number of fire tests, simulations of aircraft crash, investigations with anti tank weapons, and an explosion of a railway tank with liquid gas neighbouring a loaded CASTOR{sup R} cask. According to customer and site specific demands, different types of storage facilities are realized in Germany. Firstly, there are facilities for long-term storage, such as large ventilated central storage buildings away from reactor or ventilated storage buildings at the reactor site, ventilated underground tunnels or concrete platforms outside a building. Secondly, there are facilities for temporary storage, where casks have been positioned in horizontal orientation under a ventilated shielding cover outside a building. (authors)

  15. FFTF/IEM (Fast Flux Test Facility/Interim Examination and Maintenance) cell fuel pin weighing system

    SciTech Connect

    Gibbons, P.W.

    1987-01-01

    The Interim Examination and Maintenance (IEM) cell in the Fast Flux Test Facility (FFTF) is used for remote dissassembly of irradiated fuel and materials experiments. For those fuel experiments where the FFTF tag-gas detection system has indicated a fuel pin cladding breach, a weighing system is used in identifying that fuel pin with a reduced weight due to the escape of gaseous and volatile fission products. A fuel pin weighing machine, originally purchased for use in the Fuels and Materials Examination Facility (FMEF), was the basis for the IEM cell system. Design modifications to the original equipment were centered around adapting the machine to the differences between the two facilities and correcting deficiencies discovered during functional testing in the IEM cell mock-up.

  16. Safety issues of dry fuel storage at RSWF

    SciTech Connect

    Clarksean, R.L.; Zahn, T.P.

    1995-02-01

    Safety issues associated with the dry storage of EBR-II spent fuel are presented and discussed. The containers for the fuel have been designed to prevent a leak of fission gases to the environment. The storage system has four barriers for the fission gases. These barriers are the fuel cladding, an inner container, an outer container, and the liner at the RSWF. Analysis has shown that the probability of a leak to the environment is much less than 10{sup {minus}6} per year, indicating that such an event is not considered credible. A drop accident, excessive thermal loads, criticality, and possible failure modes of the containers are also addressed.

  17. FFTF/IEM (Fast Flux Test Facility/Interim Examination and Maintenance) cell fuel pin removal equipment

    SciTech Connect

    Greenwell, R.K.

    1987-01-01

    This paper describes a fuel pin removal device used for pin removal from irradiated fuel assemblies at the Fast Flux Test Facility (FFTF). After irradiation in the FFTF, selected fuel assemblies are remotely disassembled in the Interim Examination and Maintenance (IEM) cell. The remote disassembly, following sodium removal, consists of slitting and removing the duct and then removing the fuel pins one-at-a-time by sliding the pins from parallel attachment rails. All pins are removed from one rail before starting on the next. The new pin removal equipment has been used very successfully on the last three fuel experiments disassembled in the IEM cell, including one assembly containing residual sodium. Pin removal time has been cut in half, and this once tedious and time-consuming activity has been turned into an almost effortless evolution.

  18. Drying grain using a hydrothermally treated liquid lignite fuel

    SciTech Connect

    Bukurov, Z.; Cvijanovic, P.; Bukurov, M.; Ljubicic, B.R.

    1995-12-01

    A shortage of domestic oil and natural gas resources in Yugoslavia, particularly for agricultural and industrial purposes, has motivated the authors to explore the possibility of using liquid lignite as an alternate fuel for drying grain. This paper presents a technical and economic assessment of the possibility of retrofitting grain-drying plants currently fueled by oil or natural gas to liquid lignite fuel. All estimates are based on lignite taken from the Kovin deposit. Proposed technology includes underwater mining techniques, aqueous ash removal, hydrothermal processing, solids concentration, pipeline transport up to 120 km, and liquid lignite direct combustion. For the characterization of Kovin lignite, standard ASTM procedures were used: proximate, ultimate, ash, heating value, and Theological analyses were performed. Results from an extensive economic analysis indicate a delivered cost of US$20/ton for the liquid lignite. For the 70 of the grain-drying plants in the province of Vojvodina, this would mean a total yearly saving of about US $2,500,000. The advantages of this concept are obvious: easy to transport and store, nonflammable, nonexplosive, nontoxic, 30%-40% cheaper than imported oil and gas, domestic fuel is at hand. The authors believe that liquid lignite, rather than an alternative, is becoming more and more an imperative.

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

    SciTech Connect

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

    2013-08-01

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

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

    SciTech Connect

    Piepho, M.G.

    1998-07-20

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

  1. Combustion gas properties. 2: Natural gas fuel and dry air

    NASA Technical Reports Server (NTRS)

    Wear, J. D.; Jones, R. E.; Trout, A. M.; Mcbride, B. J.

    1985-01-01

    A series of computations has been made to produce the equilibrium temperature and gas composition for natural gas fuel and dry air. The computed tables and figures provide combustion gas property data for pressures from 0.5 to 50 atmospheres and equivalence ratios from 0 to 2.0. Only samples tables and figures are provided in this report. The complete set of tables and figures is provided on four microfiche films supplied with this report.

  2. Characterization of the radiation environment for a large-area interim spent-nuclear-fuel storage facility

    NASA Astrophysics Data System (ADS)

    Fortkamp, Jonathan C.

    Current needs in the nuclear industry and movements in the political arena indicate that authorization may soon be given for development of a federal interim storage facility for spent nuclear fuel. The initial stages of the design work have already begun within the Department of Energy and are being reviewed by the Nuclear Regulatory Commission. This dissertation addresses the radiation environment around an interim spent nuclear fuel storage facility. Specifically the dissertation characterizes the radiation dose rates around the facility based on a design basis source term, evaluates the changes in dose due to varying cask spacing configurations, and uses these results to define some applicable health physics principles for the storage facility. Results indicate that dose rates from the facility are due primarily from photons from the spent fuel and Co-60 activation in the fuel assemblies. In the modeled cask system, skyshine was a significant contribution to dose rates at distances from the cask array, but this contribution can be reduced with an alternate cask venting system. With the application of appropriate health physics principles, occupation doses can be easily maintained far below regulatory limits and maintained ALARA.

  3. Dry compliant seal for phosphoric acid fuel cell

    DOEpatents

    Granata, Jr., Samuel J.; Woodle, Boyd M.

    1990-01-01

    A dry compliant overlapping seal for a phosphoric acid fuel cell preformed f non-compliant Teflon to make an anode seal frame that encircles an anode assembly, a cathode seal frame that encircles a cathode assembly and a compliant seal frame made of expanded Teflon, generally encircling a matrix assembly. Each frame has a thickness selected to accommodate various tolerances of the fuel cell elements and are either bonded to one of the other frames or to a bipolar or end plate. One of the non-compliant frames is wider than the other frames forming an overlap of the matrix over the wider seal frame, which cooperates with electrolyte permeating the matrix to form a wet seal within the fuel cell that prevents process gases from intermixing at the periphery of the fuel cell and a dry seal surrounding the cell to keep electrolyte from the periphery thereof. The frames may be made in one piece, in L-shaped portions or in strips and have an outer perimeter which registers with the outer perimeter of bipolar or end plates to form surfaces upon which flanges of pan shaped, gas manifolds can be sealed.

  4. Utilization of polysaccharides in the drying of fuel alcohol

    SciTech Connect

    Ladisch, M.R.; Gulati, M.; Westgate, P.

    1995-12-01

    The fuel ethanol industry has grown from an annual production level of about 100 million gallons in 1978 to 1.5 billion gallons today. Technical developments which have paralleled this growth include improvements in fermentation technology, energy integration of fermentation ethanol plants, and use of improved methods of separating ethanol from water. The role of biotechnology in this expanding use of renewable resources for fuel alcohol production will be reviewed. Developments in the concentrating of fermentation ethanol by distillation, and the drying of ethanol by an adsorptive method will be presented in the context of advances in the energetics of product recovery from fermentation broths. The principles of these methods, and their current and future impact on fermentation alcohol production which uses corn will be discussed.

  5. Deposition and corrosion phenomena on aluminum surfaces under deluged dry cooling-tower condisions. Interim report

    SciTech Connect

    Wheeler, K.R.; May, R.P.; Douglas, J.G.; Tylczak, J.H.

    1981-07-01

    Deposition and corrosion on aluminum heat exchanger surfaces resulting from deluge in wet/dry cooling towers is simulated in a laboratory Corrosion/Deposition Loop (CDL). Heat exchanger deposition buildup was found to be linearly dependent on concentration factor and number of wet/dry cycles. Deionized water rising after deluge reduced rate of deposition. Laboratory data obtained from CDL relates directly to operation of the Advanced Concepts Test (ACT) demonstration cooling tower. Technology transferable to ACT shows that deposition from supersaturated solution can be effectively controlled by attention to water chemistry, pH, water conditioning, and good heat transfer design. The additional mechanism of deposition by water film evaporation is effectively managed by soft water rinsing and uniform surface wetting. Exposure of a model TRANE surface (the ACT wet/dry exchanger) produced short-term deposition extrapolating to 0.011 mm buildup in three years. Studies continue to verify 4X as maximum cycles of concentration through control of water chemistry and rinsing after deluge. Deluge water used at ACT facility is sufficiently aggressive to warrant use of Alclad to extend tube service life.

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

    SciTech Connect

    Botsch, W.; Smalian, S.; Hinterding, P.

    2013-07-01

    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

  7. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    SciTech Connect

    Wang, Lumin; Wierschke, Jonathan Brett

    2015-04-08

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H3BO3). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  8. Design review report FFTF interim storage cask

    SciTech Connect

    Scott, P.L.

    1995-01-03

    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.

  9. Research on an energy-efficient drying process. Interim report, September 24, 1982-November 30, 1983

    SciTech Connect

    Ray, R.J.

    1984-03-08

    Phase I of a program to investigate the feasibility of a membrane-based process for the recovery of energy from industrial dryer exhaust streams is covered. A Bend Research asymmetric cellulose acetate membrane was identified as being generally the most suitable membrane available for this application. Laboratory tests in a simulated drying unit demonstrated that a spiral-wound cellulose acetate membrane module is capable of recovering 40% of the water vapor in a 180/sup 0/F feed stream at 60% relative humidity while producing a permeate stream of 96 mol % water vapor. The water vapor flux through the membrane under these conditions was about 240 x 10/sup -4/cm/sup 3/ (STP)/cm/sup 2/-sec. A technical analysis indicates that this membrane-based process can be used to recover otherwise wasted heat from dryer exhaust streams. Specifically, the analysis shows that the retro-fit of this process to a typical paper dryer results in recovery of 13 times as much energy as is required to operate the necessary vacuum pumps. An economic analysis of this system indicates that, with modest improvement of membrane performance, energy recovery via this process is economically feasible. In addition, this membrane-based energy-recovery process would allow heating capacity to be added to an existing dryer without adding furnace capacity.

  10. 75 FR 23821 - Final License Renewal Interim Staff Guidance LR-ISG-2009-01: Aging Management of Spent Fuel Pool...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-04

    ... Neutron-Absorbing Materials Other Than Boraflex; Notice of Availability AGENCY: Nuclear Regulatory... Interim Staff Guidance (LR-ISG), LR-ISG-2009-01, ``Aging Management of Spent Fuel Pool Neutron-Absorbing... loss of material and loss of neutron-absorbing capability of certain nuclear power plant spent...

  11. Interim Principals.

    ERIC Educational Resources Information Center

    Beem, Kate

    2003-01-01

    An interim principal can buy a school district time to land a permanent successor. Also lists where to find an interim principal; the interim's steadying influence; Bob Wallace's wild ride as an interim principal in post-retirement; and Roger Prosise's rationale for turning to an interim appointment. (MLF)

  12. AFCI Transmutation Fuel Processes and By-Products Planning: Interim Report

    SciTech Connect

    Eric L. Shaber

    2005-09-01

    The goals of the Advanced Fuel Cycle Initiative (AFCI) Program are to reduce high-level waste volume, reduce long-lived and radiotoxic elements, and reclaim valuable energy content of spent nuclear fuel. The AFCI chartered the Fuel Development Working Group (FDWG) to develop advanced fuels in support of the AFCI goals. The FDWG organized a phased strategy of fuel development that is designed to match the needs of the AFCI program: Phase 1 - High-burnup fuels for light-water reactors (LWRs) and tri-isotopic (TRISO) fuel for gas-cooled reactors Phase 2 – Mixed oxide fuels with minor actinides for LWRs, Am transmutation targets for LWRs, inert matrix fuels for LWRs, and TRISO fuel containing Pu and other transuranium for gas-cooled reactors Phase 3 – Fertile free or low-fertile metal, ceramic, ceramic dispersed in a metal matrix (CERMET), and ceramics dispersed in a ceramic matrix (CERCER) that would be used primarily in fast reactors. Development of advanced fuels requires the fabrication, assembly, and irradiation of prototypic fuel under bounding reactor conditions. At specialized national laboratory facilities small quantities of actinides are being fabricated into such fuel for irradiation tests. Fabrication of demonstration quantities of selected fuels for qualification testing is needed but not currently feasible, because existing manual glovebox fabrication approaches result in significant radiation exposures when larger quantities of actinides are involved. The earliest demonstration test fuels needed in the AFCI program are expected to be variants of commercial mixed oxide fuel for use in an LWR as lead test assemblies. Manufacture of such test assemblies will require isolated fabrication lines at a facility not currently available in the U.S. Such facilities are now being planned as part of an Advanced Fuel Cycle Facility (AFCF). Adequate planning for and specification of actinide fuel fabrication facilities capable of producing transmutation fuels

  13. Interim Base-Level Guide for Exposure to Jet Fuel and Additives

    DTIC Science & Technology

    2011-12-01

    Equipment Personnel who may come into contact with JP-8 should wear appropriate PPE (e.g., butyl and nitrile rubber gloves, neoprene rubber headwear...applied to emerging fuels and their blends with JP-8. Contract efforts are currently underway to evaluate the toxicity of emerging fuels, and the U.S...fuels and their blends with JP-8. Contract efforts are currently underway to evaluate the toxicity of emerging fuels, and the U.S. Air Force School of

  14. Rapid determination of wood fuel moisture content using a microwave oven for drying

    SciTech Connect

    Harris, R.A.

    1982-10-01

    A method of determining moisture content (MC) of wood fuel using a microwave oven for drying the wood was evaluated by drying paired samples of five different wood fuel types in a microwave oven and a conventional oven. When compared to the conventional oven drying method, the microwave technique produces consistently lower MC determinations, although the differences are less than 1 percent. The advantage of the microwave technique is the speed at which MC determinations can be determined (less than 15 minutes). Schedules for drying five wood fuel types are presented. (Refs. 7).

  15. Status of work at PNL supporting dry storage of spent fuel

    SciTech Connect

    Cunningham, M.E.; McKinnon, M.A.; Michener, T.E.; Thomas, L.E.; Thornhill, C.K.

    1993-01-01

    This report discusses three projects related to dry storage of light-water reactor spent fuel which are being conducted at Pacific Northwest Laboratory. Performance testing of six dry storage systems (four metal casks and two concrete storage systems) has been completed and results compiled. Two computer codes for predicting spent fuel and storage system thermal performance, COBRA-SFS and HYDRA-II, have been developed and have been reviewed by the US Nuclear Regulatory Commission. Air oxidation testing of spent fuel was conducted from 1984 through 1990 to obtain data to support recommendations of temperature-time limits for air dry storage of spent light-water reactor fuel.

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

    SciTech Connect

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

    1992-04-01

    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.

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

    SciTech Connect

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

    1992-04-01

    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.

  18. Solar hydrogen production: renewable hydrogen production by dry fuel reforming

    NASA Astrophysics Data System (ADS)

    Bakos, Jamie; Miyamoto, Henry K.

    2006-09-01

    SHEC LABS - Solar Hydrogen Energy Corporation constructed a pilot-plant to demonstrate a Dry Fuel Reforming (DFR) system that is heated primarily by sunlight focusing-mirrors. The pilot-plant consists of: 1) a solar mirror array and solar concentrator and shutter system; and 2) two thermo-catalytic reactors to convert Methane, Carbon Dioxide, and Water into Hydrogen. Results from the pilot study show that solar Hydrogen generation is feasible and cost-competitive with traditional Hydrogen production. More than 95% of Hydrogen commercially produced today is by the Steam Methane Reformation (SMR) of natural gas, a process that liberates Carbon Dioxide to the atmosphere. The SMR process provides a net energy loss of 30 to 35% when converting from Methane to Hydrogen. Solar Hydrogen production provides a 14% net energy gain when converting Methane into Hydrogen since the energy used to drive the process is from the sun. The environmental benefits of generating Hydrogen using renewable energy include significant greenhouse gas and criteria air contaminant reductions.

  19. Research on fire-resistant diesel fuel. Interim report 1 Oct 79-31 Dec 81

    SciTech Connect

    Weatherford, W.D. Jr; Fodor, G.E.; Kanakia, M.D.; Naegeli, D.W.; Wright, B.R.

    1981-12-01

    When development of aqueous fire-resistant diesel fuel (FRF) was previously reported, it was shown that clear-to-hazy water-in-fuel, diesel fuel micro-emulsions could be prepared and that they exhibit reduced mist flammability and self-extinguishing pool fires at temperatures above the base fuel flash point. It was also demonstrated that unmodified diesel engines start, idle, and run without difficulty on such fuels. Research has been continued to establish compositional requirements for base fuels, surfactants, and water used in FRF formulations. DF-2, DF-1, DF-A, and NATO diesel fuel samples were obtained from refineries, bulk storage, and service stations. Aromatic concentrate (AC) products from various sources were evaluated for use in adjusting the total aromatic ring carbon (TARC) content of FRF formulations. Neat base fuel and AC-containing base fuel TARC effects on microemulsification efficacy were established for water containing various amounts of total dissolved solids and for the amide/amine/soap emulsifier with various levels of total acid number.

  20. 76 FR 9381 - Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-02-17

    ... Gordon, Structural Mechanics and Materials Branch, Division of Spent Fuel Storage and Transportation... ISG-23 should be directed to Matthew Gordon, Structural Mechanics and Materials Branch, Division of.... Michele Sampson, Acting Chief, Structural Mechanics and Materials Branch, Division of Spent Fuel...

  1. Dry hog fuel to improve effiency, cut emissions

    SciTech Connect

    Schwieger, B.

    1980-02-01

    Various dryers in wood-fired powerplants are described and it is stated that when moisture levels of hog fuel rise above 55%, boilers cannot produce enough heat to sustain combustion. Methods to avoid this problem are suggested and include the burning of a low-moisture fuel in conjunction with the hog fuel and the installation of a dryer to remove some moisture from the fuel before it enter the furnace. It is generally agreed that flue-gas dryers should be considered in the design of hog-fuel-fired steam sytems whenever fuel moisture exceeds about 50%.

  2. Status of work at PNL supporting dry storage of spent fuel

    SciTech Connect

    Cunningham, M.E.; McKinnon, M.A.; Michener, T.E.; Thomas, L.E.; Thornhill, C.K.

    1992-01-01

    Three projects related to dry storage of light-water reactor spent fuel are being conducted at Pacific Northwest Laboratory. Performance testing of six dry storage systems (four metal casks and two concrete storage systems) has been completed and results compiled. Two computer codes for predicting spent fuel and storage system thermal performance, COBRA-SFS and HYDRA-II, have been developed and have been reviewed by the US Nuclear Regulatory Commission. Air oxidation testing of spent fuel was conducted from 1984 through 1990 to obtain data to support recommendations of temperature-time limits for air dry storage for periods up to 40 years.

  3. Alternative fuel vehicles: The emerging emissions picture. Interim results, Summer 1996

    SciTech Connect

    1996-10-01

    In this pamphlet, program goal, description, vehicles/fuels tested, and selected emissions results are given for light-duty and heavy-duty vehicles. Other NREL R&D programs and publications are mentioned briefly.

  4. Drying results of K-Basin fuel element 0309M (Run 3)

    SciTech Connect

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

    1998-07-01

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

    SciTech Connect

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

    1997-06-01

    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.

  6. Advanced Fuels for LWRs: Fully-Ceramic Microencapsulated and Related Concepts FY 2012 Interim Report

    SciTech Connect

    R. Sonat Sen; Brian Boer; John D. Bess; Michael A. Pope; Abderrafi M. Ougouag

    2012-03-01

    This report summarizes the progress in the Deep Burn project at Idaho National Laboratory during the first half of fiscal year 2012 (FY2012). The current focus of this work is on Fully-Ceramic Microencapsulated (FCM) fuel containing low-enriched uranium (LEU) uranium nitride (UN) fuel kernels. UO2 fuel kernels have not been ruled out, and will be examined as later work in FY2012. Reactor physics calculations confirmed that the FCM fuel containing 500 mm diameter kernels of UN fuel has positive MTC with a conventional fuel pellet radius of 4.1 mm. The methodology was put into place and validated against MCNP to perform whole-core calculations using DONJON, which can interpolate cross sections from a library generated using DRAGON. Comparisons to MCNP were performed on the whole core to confirm the accuracy of the DRAGON/DONJON schemes. A thermal fluid coupling scheme was also developed and implemented with DONJON. This is currently able to iterate between diffusion calculations and thermal fluid calculations in order to update fuel temperatures and cross sections in whole-core calculations. Now that the DRAGON/DONJON calculation capability is in place and has been validated against MCNP results, and a thermal-hydraulic capability has been implemented in the DONJON methodology, the work will proceed to more realistic reactor calculations. MTC calculations at the lattice level without the correct burnable poison are inadequate to guarantee zero or negative values in a realistic mode of operation. Using the DONJON calculation methodology described in this report, a startup core with enrichment zoning and burnable poisons will be designed. Larger fuel pins will be evaluated for their ability to (1) alleviate the problem of positive MTC and (2) increase reactivity-limited burnup. Once the critical boron concentration of the startup core is determined, MTC will be calculated to verify a non-positive value. If the value is positive, the design will be changed to require

  7. Interim report re: component parts for proton-exchange membrane fuel cells

    SciTech Connect

    George Marchetti

    1999-10-01

    The purpose of the first phase of the grant project is to design, develop and test a simplified fuel cell electrode structure for use in proton-exchange membrane fuel cells (''PEMFC''). By simplifying the structure of the electrode, mass production manufacturing efficiencies can be brought into play which will result in significant cost reductions for this fuel cell component. With a reduction in the cost of this key fuel cell component overall costs for PEMFC's can be brought within the commercialization target range of about US$100 per kilowatt for the fuel cell stack. Fuel cell electrodes are necessarily ''multi-layered'' composites. Multi-layers are required because of the several functions that the electrode must be able to perform in the working PEM fuel cell. The current generation of state-of-the-art porous fuel cell electrodes for PEMFC's is comprised of three primary layers. The first layer is the catalyst layer. Since hydrogen is the fuel used in this project and air is used as the oxidant, the catalyst must be capable of adsorbing hydrogen and oxygen from the air. While work is constantly on-going with respect to new hydrogen or oxygen catalysts, the best available catalyst at present for both of the reactant gases is platinum. To be effective, the catalyst (1) must be exposed to a constant flow of the respective reactant gas; (2) must be in intimate contact with the proton-exchange membrane; and (3) must be a finely divided catalyst and have a large specific surface area, especially on the oxidant side where the electrochemical reaction is slower by several orders of magnitude. The second layer is the substrate layer. The substrate layer provides structural support for the finely divided catalyst. It also functions as an electronic junction for conducting electricity produced by the electrochemical reaction from the catalyst layer to the bipolar plate of the fuel cell. In state-of-the-art PEMFC's, this layer is comprised of carbon particles (onto which

  8. Extinguishing in-flight engine fuel-leak fires with dry chemicals

    NASA Technical Reports Server (NTRS)

    Altman, R. L.

    1981-01-01

    The fire extinguishant storage temperature requirements were examined for several commercially available dry chemicals. Particular emphasis was placed on the development of dry powder extinguishant that, when discharged into a jet engine fuel leak fire, would stick to the hot surfaces. Moreover, after putting out the initial fire, these extinguishants would act as antireignition catalysts, even when the fuel continued to leak onto the heated surface.

  9. Statistical analysis of oxidation rates for K Basin fuel in dry air

    SciTech Connect

    Trimble, D.J.

    1998-02-06

    Test data from oxidation of K Basin fuel (SNF) samples in dry air were reviewed, and linear reaction rates were derived on a time-average basis. The derived rates were compared to literature data for unirradiated uranium in dry air using rate law of the form log(rate) = a + b (I/T). The analyses found differences between the SNF data and the literature data. Oxidation rate below 150 C was higher for K Basin fuel than for unirradiated uranium.

  10. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    SciTech Connect

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

    2008-03-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

  11. Sensitivity analysis of a dry-processed Candu fuel pellet's design parameters

    SciTech Connect

    Choi, Hangbok; Ryu, Ho Jin

    2007-07-01

    Sensitivity analysis was carried out in order to investigate the effect of a fuel pellet's design parameters on the performance of a dry-processed Canada deuterium uranium (CANDU) fuel and to suggest the optimum design modifications. Under a normal operating condition, a dry-processed fuel has a higher internal pressure and plastic strain due to a higher fuel centerline temperature when compared with a standard natural uranium CANDU fuel. Under a condition that the fuel bundle dimensions do not change, sensitivity calculations were performed on a fuel's design parameters such as the axial gap, dish depth, gap clearance and plenum volume. The results showed that the internal pressure and plastic strain of the cladding were most effectively reduced if a fuel's element plenum volume was increased. More specifically, the internal pressure and plastic strain of the dry-processed fuel satisfied the design limits of a standard CANDU fuel when the plenum volume was increased by one half a pellet, 0.5 mm{sup 3}/K. (authors)

  12. New co-products from grain-based fuel ethanol production and their drying performance

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Fuel ethanol production in the U.S. and elsewhere is an important and growing industry. In the U.S, about 40% of annual corn production is now converted into fuel ethanol. During co-product recovery, condensed distillers solubles (CDS) has to be mixed with distillers wet grains before drying due to ...

  13. Diesel Emission Control -- Sulfur Effects (DECSE) Program; Phase I Interim Date Report No. 3: Diesel Fuel Sulfur Effects on Particulate Matter Emissions

    SciTech Connect

    DOE; ORNL; NREL; EMA; MECA

    1999-11-15

    The Diesel Emission Control-Sulfur Effects (DECSE) is a joint government/industry program to determine the impact of diesel fuel sulfur levels on emission control systems whose use could lower emissions of nitrogen oxides (NO{sub x}) and particulate matter (PM) from on-highway trucks in the 2002--2004 model years. Phase 1 of the program was developed with the following objectives in mind: (1) evaluate the effects of varying the level of sulfur content in the fuel on the emission reduction performance of four emission control technologies; and (2) measure and compare the effects of up to 250 hours of aging on selected devices for multiple levels of fuel sulfur content. This interim report covers the effects of diesel fuel sulfur level on particulate matter emissions for four technologies.

  14. Method for producing a dried coal fuel having a reduced tendency to spontaneously ignite from a low rank coal

    SciTech Connect

    Li, Y.H.; Bonnecaze, B.F.; Matthews, J.D.; Skinner, J.L.; Wunderlich, D.K.

    1983-08-02

    A method is disclosed for producing a dried coal fuel having a reduced tendency to spontaneously ignite from a low rank coal by drying the low rank coal and thereafter cooling the dried coal to a temperature below about 100/sup 0/F. Optionally the dried coal is partially oxidized prior to cooling and optionally the dried coal is mixed with a deactivating fluid.

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

    SciTech Connect

    CARLSON, A.B.

    1999-02-24

    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.

  16. Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

    SciTech Connect

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1982-04-30

    The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test.

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

    SciTech Connect

    Bailey, W.J.

    1990-02-01

    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.

  18. Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

    SciTech Connect

    Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; Poulson, Daniel; Bacon, Jeffrey Darnell; Chichester, David; Fabritius, Joseph; Fellows, Shelby; Plaud-Ramos, Kenie Omar; Morley, Deborah Jean; Winston, Philip

    2016-04-29

    In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may find other safety and security or safeguards applications, such as arms control verification.

  19. Spent nuclear fuel vacuum drying thermal-hydraulic analysis and dynamic model development status report

    SciTech Connect

    Irwin, J.J.; Ogden, D.M., Westinghouse Hanford

    1996-08-28

    This report summarizes preliminary thermal hydraulic scoping analysis and model development associated with the K Basin spent fuel MCO draining and vacuum drying system. The purpose of the draining and drying system is to remove all free water from the interior of the MCO, baskets, and fuel prior to back filling with inert gas and transfer to the hot conditioning process. Dominant physical processes and parameters are delineated and related quantitatively. Minimum dynamic modeling capability required to simulate the process of transporting heat to the residual water on the fuel and transport of the steam produced from the system by vacuum pumping are defined.

  20. Interim Expertise

    ERIC Educational Resources Information Center

    Anyaso, Hilary Hurd

    2009-01-01

    The Registry for College and University Presidents places former executives in interim presidential and other senior-level posts and is familiar with the challenges interim executives and institutions encounter in times of leadership transitions. However, the one big advantage interims bring to institutions, says Registry Vice President Kevin J.…

  1. Spent fuel behavior under abnormal thermal transients during dry storage

    SciTech Connect

    Stahl, D.; Landow, M.P.; Burian, R.J.; Pasupathi, V.

    1986-01-01

    This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment was heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.

  2. Extinction of in-flight engine fuel-leak fires with dry chemicals

    NASA Technical Reports Server (NTRS)

    Altman, R. L.

    1983-01-01

    The dry chemicals discussed here are seen as having a greater weight effectiveness than the halons in current use for controlling fuel-leak fires, especially in the presence of high airflow rates. The commercial dry chemicals K2CO3, KHCO3, and KC2N2H3O3 are found to be more effective than CF2ClBr and CF3Br in delaying the hot-surface reignition of fuel-leak fires after initial extinguishment. Experimental dry chemical formulations of potassium dawsonite, KAl(OH)2CO3, and of KCl and KI are seen as being even more weight effective than the above-mentioned commercial dry chemicals. It is noted, however, that the suitability and effectiveness of dry chemicals in controlling engine nacele fires has not yet been demonstrated in test aircraft.

  3. Spent nuclear fuel project cold vacuum drying facility operations manual

    SciTech Connect

    IRWIN, J.J.

    1999-05-12

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  4. Equipment concepts for dry intercask transfer of spent fuel

    SciTech Connect

    Schneider, K.J.

    1983-07-01

    This report documents the results of a study of preconceptual design and analysis of four intercask transfer concepts. The four concepts are: a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); a small hot cell containing equipment for transferring fuel between shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept); and a large hot cell, shielded by an earthen berm, that houses equipment for handling fuel between casks that enter and leave the cell on a single cart (igloo concept). The casks considered in this study are most of the transport casks currently operable in the USA, and the storage casks designated REA-2023 and GNS Castor-V. Exclusive of basic services assumed to be provided at the host site, the design and capital costs are estimated to range from $9 to $13 million. The portion of capital costs for portable equipment (for potential later use at another site) was estimated to range from 70% to 98%, depending on the concept. Increasing portability from a range of 70 to 90% to 98% adds $2 to 4 million to the capital costs. Operating costs are estimated at about $2 million/year for all concepts. Implementation times range from about 18 months for the more conventional systems to 40 months for the more unique systems. Times and costs for relocation to another site are 10 to 14 months and about $1 million, plus shipping costs and costs of new construction at the new site. All concepts have estimated capacities for fuel transfer at least equal to the criterion set for this study. Only the hot cell concepts have capability for recanning or repair of canisters. Some development is believed to be required for the turntable and shuttle concepts, but none for the other two concepts.

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

    SciTech Connect

    Palmer, A.J.

    1995-12-31

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

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

    SciTech Connect

    CARRELL, R D

    2002-07-16

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

  7. Gaseous fuels production from dried sewage sludge via air gasification.

    PubMed

    Werle, Sebastian; Dudziak, Mariusz

    2014-07-01

    Gasification is a perspective alternative method of dried sewage sludge thermal treatment. For the purpose of experimental investigations, a laboratory fixed-bed gasifier installation was designed and built. Two sewage sludge (SS) feedstocks, taken from two typical Polish wastewater treatment systems, were analysed: SS1, from a mechanical-biological wastewater treatment system with anaerobic stabilization (fermentation) and high temperature drying; and (SS2) from a mechanical-biological-chemical wastewater treatment system with fermentation and low temperature drying. The gasification results show that greater oxygen content in sewage sludge has a strong influence on the properties of the produced gas. Increasing the air flow caused a decrease in the heating value of the produced gas. Higher hydrogen content in the sewage sludge (from SS1) affected the produced gas composition, which was characterized by high concentrations of combustible components. In the case of the SS1 gasification, ash, charcoal, and tar were produced as byproducts. In the case of SS2 gasification, only ash and tar were produced. SS1 and solid byproducts from its gasification (ash and charcoal) were characterized by lower toxicity in comparison to SS2. However, in all analysed cases, tar samples were toxic.

  8. Technical Issues and Characterization for Fuel and Sludge in Hanford K Basins

    SciTech Connect

    MAKENAS, B.J.

    2000-06-01

    Technical Issues for the interim dry storage of N Reactor Spent Nuclear Fuel (SNF) are discussed. Characterization data from fuel, to support resolution of these issues, are reviewed and new results for the oxidation of fuel in a moist atmosphere and the drying of whole fuel elements are presented. Characterization of associated K basin sludge is also discussed in light of a newly adopted disposal pathway.

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

    SciTech Connect

    1996-07-01

    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.

  10. Operation of polymer electrolyte membrane fuel cells with dry feeds: Design and operating strategies

    NASA Astrophysics Data System (ADS)

    Hogarth, Warren H. J.; Benziger, Jay B.

    The operation of polymer electrolyte membrane fuel cells (PEMFCs) with dry feeds has been examined with different fuel cell flow channel designs as functions of pressure, temperature and flow rate. Auto-humidified (or self-humidifying) PEMFC operation is improved at higher pressures and low gas velocities where axial dispersion enhances "back-mixing" of the product water with the dry feed. We demonstrate auto-humidified operation of the channel-less, self-draining fuel cell, based on a stirred tank reactor; data is presented showing auto-humidified operation from 25 to 115 °C at 1 and 3 atm. Design and operating requirements are derived for the auto-humidified operation of the channel-less, self-draining fuel cell. The auto-humidified self-draining fuel cell outperforms a fully humidified serpentine flow channel fuel cell at high current densities. The new design offers substantial benefits for simplicity of operation and control including: the ability to self-drain reducing flooding, the ability to uniformly disperse water removing current gradients and the ability to operate on dry feeds eliminating the need for humidifiers. Additionally, the design lends itself well to a modular design concept.

  11. Volatile organic compound emissions from dry mill fuel ethanol production.

    PubMed

    Brady, Daniel; Pratt, Gregory C

    2007-09-01

    Ethanol fuel production is growing rapidly in the rural Midwest, and this growth presents potential environmental impacts. In 2002, the U.S. Environmental Protection Agency (EPA) and the Minnesota Pollution Control Agency (MPCA) entered into enforcement actions with 12 fuel ethanol plants in Minnesota. The enforcement actions uncovered underreported emissions and resulted in consent decrees that required pollution control equipment be installed. A key component of the consent decrees was a requirement to conduct emissions tests for volatile organic compounds (VOCs) with the goal of improving the characterization and control of emissions. The conventional VOC stack test method was thought to underquantify total VOC emissions from ethanol plants. A hybrid test method was also developed that involved quantification of individual VOC species. The resulting database of total and speciated VOC emissions from 10 fuel ethanol plants is relatively small, but it is the most extensive to date and has been used to develop and gauge compliance with permit limits and to estimate health risks in Minnesota. Emissions were highly variable among facilities and emissions units. In addition to the variability, the small number of samples and the presence of many values below detection limits complicate the analysis of the data. To account for these issues, a nested bootstrap procedure on the Kaplan-Meier method was used to calculate means and upper confidence limits. In general, the fermentation scrubbers and fluid bed coolers emitted the largest mass of VOC emissions. Across most facilities and emissions units ethanol was the pollutant emitted at the highest rate. Acetaldehyde, acetic acid, and ethyl acetate were also important emissions from some units. Emissions of total VOCs, ethanol, and some other species appeared to be a function of the beer feed rate, although the relationship was not reliable enough to develop a production rate-based emissions factor.

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

    SciTech Connect

    Bryan, Charles R.; Enos, David George

    2014-07-01

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

  13. Progress and interim results of the INPRO joint study on assessment of INS based on closed nuclear fuel cycle with fast reactors

    SciTech Connect

    Usanov, Vladimir; Raj, Baldev; Vasile, Alfredo

    2007-07-01

    The purpose of the work is to review interim results of the Joint Study on assessment of an Innovative Nuclear System based on a Closed Nuclear Fuel Cycle with Fast Reactors (INS CNFC-FR). This study is a part of the IAEA international project for innovative reactors and fuel cycle technologies (INPRO). Now it is being implemented by Canada, China, France, India, Japan, Republic of Korea, Russia, and Ukraine. A report on results of implementation of the first phase of the Joint Study was presented to the INPRO Steering Committee meeting in December 2006. It was also agreed by the Joint Study participants to reveal these results to broader discussion at scientific conferences and meetings. The authors' interpretation of the Joint Study findings and issues is presented in the paper. (authors)

  14. Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI)

    SciTech Connect

    Trond Bjornard; Philip C. Durst

    2012-05-01

    This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA) of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a country’s safeguards agreement with

  15. Imaging Spent Fuel in Dry Storage Casks with Cosmic Ray Muons

    SciTech Connect

    Durham, J. Matthew; Dougan, Arden

    2015-11-05

    Highly energetic cosmic ray muons are a natural source of ionizing radiation that can be used to make tomographic images of the interior of dense objects. Muons are capable of penetrating large amounts of shielding that defeats typical radiographic probes like neutrons or photons. This is the only technique which can examine spent nuclear fuel rods sealed inside dry casks.

  16. Combustion Gas Properties I-ASTM Jet a Fuel and Dry Air

    NASA Technical Reports Server (NTRS)

    Jones, R. E.; Trout, A. M.; Wear, J. D.; Mcbride, B. J.

    1984-01-01

    A series of computations was made to produce the equilibrium temperature and gas composition for ASTM jet A fuel and dry air. The computed tables and figures provide combustion gas property data for pressures from 0.5 to 50 atmospheres and equivalence ratios from 0 to 2.0.

  17. Combustion characteristics of dry coal-powder-fueled adiabatic diesel engine: Final report

    SciTech Connect

    Kakwani, R.M.; Kamo, R.

    1989-01-01

    This report describes the progress and findings of a research program aimed at investigating the combustion characteristics of dry coal powder fueled diesel engine. During this program, significant achievements were made in overcoming many problems facing the coal-powder-fueled engine. The Thermal Ignition Combustion System (TICS) concept was used to enhance the combustion of coal powder fuel. The major coal-fueled engine test results and accomplishments are as follows: design, fabrication and engine testing of improved coal feed system for fumigation of coal powder to the intake air; design, fabrication and engine testing of the TICS chamber made from a superalloy material (Hastelloy X); design, fabrication and engine testing of wear resistant chrome oxide ceramic coated piston rings and cylinder liner; lubrication system was improved to separate coal particles from the contaminated lubricating oil; control of the ignition timing of fumigated coal powder by utilizing exhaust gas recirculation (EGR) and variable TICS chamber temperature; coal-fueled engine testing was conducted in two configurations: dual fuel (with diesel pilot) and 100% coal-fueled engine without diesel pilot or heated intake air; cold starting of the 100% coal-powder-fueled engine with a glow plug; and coal-fueled-engine was operated from 800 to 1800 rpm speed and idle to full load engine conditions.

  18. Cost Sensitivity Analysis for Consolidated Interim Storage of Spent Fuel: Evaluating the Effect of Economic Environment Parameters

    SciTech Connect

    Cumberland, Riley M.; Williams, Kent Alan; Jarrell, Joshua J.; Joseph, III, Robert Anthony

    2016-12-01

    This report evaluates how the economic environment (i.e., discount rate, inflation rate, escalation rate) can impact previously estimated differences in lifecycle costs between an integrated waste management system with an interim storage facility (ISF) and a similar system without an ISF.

  19. Dry Bag Isostatic Pressing for Improved Green Strength of Nuclear Fuel Pellets

    SciTech Connect

    G. W. Egeland; L. D. Zuck; W. R. Cannon; P. A. Lessing; P. G. Medvedev

    2010-11-01

    Dry bag isostatic pressing is proposed for mass production of nuclear fuel pellets. Dry bag isostatically pressed rods of a fuel surrogate (95% CeO2-5% HfO2) 200 mm long by 8 mm diameter were cut into pellets using a wire saw. Four different binder and two different CeO2 powder sources were investigated. The strength of the isostatically pressed pellets for all binder systems measured by diametral compression was about 50% higher than pellets produced by uniaxial dry pressing at the same pressure. It was proposed that the less uniform density of uniaxially pressed pellets accounted for the lower strength. The strength of pellets containing CeO2 powder with significantly higher moisture content was five times higher than pellets containing CeO2 powder with a low moisture content. Capillary pressure of the moisture was thought to supply the added binding strength.

  20. Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

    DOE PAGES

    Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; ...

    2016-04-29

    In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may findmore » other safety and security or safeguards applications, such as arms control verification.« less

  1. Life cycle assessment of fuel ethanol derived from corn grain via dry milling.

    PubMed

    Kim, Seungdo; Dale, Bruce E

    2008-08-01

    Life cycle analysis enables to investigate environmental performance of fuel ethanol used in an E10 fueled compact passenger vehicle. Ethanol is derived from corn grain via dry milling. This type of analysis is an important component for identifying practices that will help to ensure that a renewable fuel, such as ethanol, may be produced in a sustainable manner. Based on data from eight counties in seven Corn Belt states as corn farming sites, we show ethanol derived from corn grain as E10 fuel would reduce nonrenewable energy and greenhouse gas emissions, but would increase acidification, eutrophication and photochemical smog, compared to using gasoline as liquid fuel. The ethanol fuel systems considered in this study offer economic benefits, namely more money returned to society than the investment for producing ethanol. The environmental performance of ethanol fuel system varies significantly with corn farming sites because of different crop management practices, soil properties, and climatic conditions. The dominant factor determining most environmental impacts considered here (i.e., greenhouse gas emissions, acidification, eutrophication, and photochemical smog formation) is soil related nitrogen losses (e.g., N2O, NOx, and NO3-). The sources of soil nitrogen include nitrogen fertilizer, crop residues, and air deposition. Nitrogen fertilizer is probably the primary source. Simulations using an agro-ecosystem model predict that planting winter cover crops would reduce soil nitrogen losses and increase soil organic carbon levels, thereby greatly improving the environmental performance of the ethanol fuel system.

  2. Drying results of K-Basin fuel element 3128W (run 2)

    SciTech Connect

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

    1998-07-01

    An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-East 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 N-Reactor 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 an open K-East canister (3128W) during the first fuel selection campaign conducted in 1995, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. Although it was judged to be breached during in-basin (i.e., K-Basin) examinations, visual inspection of this fuel element in the hot cell indicated that it was likely intact. Some scratches on the coating covering the cladding were identified before the furnace test. The drying test was conducted in the Whole Element Furnace Testing System located in G-Cell within the PTL. This test system is composed of three basic systems: the in-cell furnace equipment, the system gas loop, and the analytical instrument package. Element 3128W was subjected to the drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step. Results of the Pressure Rise and Gas Evolution Tests suggest that most of the free water in the system was released during the extended CVD cycle (68 hr versus 8 hr for the first run). An additional {approximately}0.34 g of water was released during the subsequent HVD phase, characterized by multiple water release peaks, with a principle peak at {approximately}180 C. This additional water is attributed to decomposition of a uranium hydrate (UO{sub 4}{center_dot}4H{sub 2}O/UO{sub 4}{center_dot}2H{sub 2}O) coating that was observed to be covering the surface

  3. Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels.

    SciTech Connect

    Wolf, S. F.

    1999-03-24

    The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns.

  4. Use of filler materials to aid spent nuclear fuel dry storage

    SciTech Connect

    Anderson, K.J.

    1981-09-01

    The use of filler materials (also known as stabilizer or encapsulating materials) was investigated in conjunction with the dry storage of irradiated light water reactor (LWR) fuel. The results of this investigation appear to be equally valid for the wet storage of fuel. The need for encapsulation and suitable techniques for closing was also investigated. Various materials were reviewed (including solids, liquids, and gases) which were assumed to fill the void areas within a storage can containing either intact or disassembled spent fuel. Materials were reviewed and compared on the basis of cost, thermal characteristics, and overall suitability in the proposed environment. A thermal analysis was conducted to yield maximum centerline and surface temperatures of a design basis fuel encapsulated within various filler materials. In general, air was found to be the most likely choice as a filler material for the dry storage of spent fuel. The choice of any other filler material would probably be based on a desire, or need, to maximize specific selection criteria, such as surface temperatures, criticality safety, or confinement.

  5. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    DOE PAGES

    Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena; ...

    2017-10-22

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

  6. Drying of Floodplain Forests Associated with Water-Level Decline in the Apalachicola River, Florida - Interim Results, 2006

    USGS Publications Warehouse

    Darst, Melanie R.; Light, Helen M.

    2007-01-01

    Floodplain forests of the Apalachicola River, Florida, are drier in composition today (2006) than they were before 1954, and drying is expected to continue for at least the next 50 years. Drier forest composition is probably caused by water-level declines that occurred as a result of physical changes in the main channel after 1954 and decreased flows in spring and summer months since the 1970s. Forest plots sampled from 2004 to 2006 were compared to forests sampled in the late 1970s (1976-79) using a Floodplain Index (FI) based on species dominance weighted by the Floodplain Species Category, a value that represents the tolerance of tree species to inundation and saturation in the floodplain and consequently, the typical historic floodplain habitat for that species. Two types of analyses were used to determine forest changes over time: replicate plot analysis comparing present (2004-06) canopy composition to late 1970s canopy composition at the same locations, and analyses comparing the composition of size classes of trees on plots in late 1970s and in present forests. An example of a size class analysis would be a comparison of the composition of the entire canopy (all trees greater than 7.5 cm (centimeter) diameter at breast height (dbh)) to the composition of the large canopy tree size class (greater than or equal to 25 cm dbh) at one location. The entire canopy, which has a mixture of both young and old trees, is probably indicative of more recent hydrologic conditions than the large canopy, which is assumed to have fewer young trees. Change in forest composition from the pre-1954 period to approximately 2050 was estimated by combining results from three analyses. The composition of pre-1954 forests was represented by the large canopy size class sampled in the late 1970s. The average FI for canopy trees was 3.0 percent drier than the average FI for the large canopy tree size class, indicating that the late 1970s forests were 3.0 percent drier than pre-1954

  7. Optimization of a Dry, Mixed Nuclear Fuel Storage Array for Nuclear Criticality Safety

    NASA Astrophysics Data System (ADS)

    Baranko, Benjamin T.

    A dry storage array of used nuclear fuel at the Idaho National Laboratory contains a mixture of more than twenty different research and test reactor fuel types in up to 636 fuel storage canisters. New analysis demonstrates that the current arrangement of the different fuel-type canisters does not minimize the system neutron multiplication factor (keff), and that the entire facility storage capacity cannot be utilized without exceeding the subcritical limit (ksafe) for ensuring nuclear criticality safety. This work determines a more optimal arrangement of the stored fuels with a goal to minimize the system keff, but with a minimum of potential fuel canister relocation movements. The solution to this multiple-objective optimization problem will allow for both an improvement in the facility utilization while also offering an enhancement in the safety margin. The solution method applies stochastic approximation and a Tabu search metaheuristic to an empirical model developed from supporting MCNP calculations. The results establish an optimal relocation of between four to sixty canisters, which will allow the current thirty-one empty canisters to be used for storage while reducing the array keff by up to 0.018 +/- 0.003 relative to the current arrangement.

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

    SciTech Connect

    IRWIN, J.J.

    1999-07-02

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

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

    SciTech Connect

    IRWIN, J.J.

    2000-02-03

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  10. Dry, portable calorimeter for nondestructive measurement of the activity of nuclear fuel

    DOEpatents

    Beyer, Norman S.; Lewis, Robert N.; Perry, Ronald B.

    1976-01-01

    The activity of a quantity of heat-producing nuclear fuel is measured rapidly, accurately and nondestructively by a portable dry calorimeter comprising a preheater, an array of temperature-controlled structures comprising a thermally guarded temperature-controlled oven, and a calculation and control unit. The difference between the amounts of electric power required to maintain the oven temperature with and without nuclear fuel in the oven is measured to determine the power produced by radioactive disintegration and hence the activity of the fuel. A portion of the electronic control system is designed to terminate a continuing sequence of measurements when the standard deviation of the variations of the amount of electric power required to maintain oven temperature is within a predetermined value.

  11. Ceramic anode catalyst for dry methane type molten carbonate fuel cell

    NASA Astrophysics Data System (ADS)

    Tagawa, T.; Yanase, A.; Goto, S.; Yamaguchi, M.; Kondo, M.

    Oxide catalyst materials for methane oxidation were examined in order to develop the anode electrode for molten carbonate type fuel cell (MCFC). As a primary selection, oxides such as lanthanum (La 2O 3) and samarium (Sm 2O 3) were selected from screening experiments of TPD, TG and tubular reactor. Composite materials of these oxides with titanium fine powder were assembled into a cell unit for MCFC as the anode electrode. Steady-state activities were observed with these anode electrode materials when hydrogen was used as a fuel. When methane was directly charged to anode as a fuel (dry methane operation), a power generation with steady state was observed on both lanthanum and samarium composites after gradual decrease of open circuit electromotive force (OCV) and closed circuit current (CCI). The steady-state activity held as long as 144 h of continuous operation.

  12. Spent fuel dry storage technology development: Report of consolidated thermal data

    NASA Astrophysics Data System (ADS)

    Lundberg, W. L.

    1980-09-01

    A drywell/sealed cask technique for spent fuel storage is discussed. Experiments indicate that PWR fuel with decay heat levels in excess of 2 kW could be stored in isolated drywells in Nevada test site soil without exceeding the current fuel clad temperature limit (715 F). The ability to thermally analyze near surface drywells and above ground storage casks is assessed. It is concluded that the required analysis procedures, computer programs, etc., are already developed and available. Soil thermal conductivity requires additional study to better understand the soil drying mechanism and effects of moisture. Work is also required to develop an internal canister subchannel model. In addition, the ability of the overall drywell thermal model to accommodate thermal interaction effects between adjacent drywells should be confirmed.

  13. Development of accelerated fuel-engines qualification procedures methodology. Volume II. Appendices. Interim report for Oct 80-Sep 81

    SciTech Connect

    Russell, J.A.; Cuellar, J.P. Jr; Tyler, J.C.; Erwin, J.; Knutson, W.K.

    1981-12-01

    This volume of appendices includes: AVRADCOM Aircraft/Engine Component Listings; AFLRL Elastomer Swell and Hardness Data; AFLRL Elastomer Retention Properties; JFC 78 500-Hour Shale Oil Fuel Plan of Test, April 20, 1981; JFC 78 Shale Oil Fuel Plan of Test, Revised July 30, 1981; Pre and Post Test Definition Data; Spot Calibration; Flow Test; Fuel Characteristics and Lubricity and Interfacial Tension; Detailed Component Test Results; Disassembly and Inspection Results; Identification of Critical Fuel System Components; Existing Engine Qualifications Procedures; NATO Standard Engine Laboratory Test-Edition June 80; Fuel System Components Qualification Procedures; and EPA Heavy-Duty Diesel Engine Exhaust Emissions Certification and Test Procedures.

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

    SciTech Connect

    IRWIN, J.J.

    2000-11-18

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

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

    SciTech Connect

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

    2003-08-17

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

    SciTech Connect

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

    2003-03-17

    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.

  17. Performance of Trasuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Interim Report, Including Void Reactivity Evaluation

    SciTech Connect

    Michael A. Pope; Brian Boer; Gilles Youinou; Abderrafi M. Ougouag

    2011-03-01

    The current focus of the Deep Burn Project is on once-through burning of transuranice (TRU) in light water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles would be pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell calculations have been performed using the DRAGON-4 code in order assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells containing typical UO2 and MOX fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Loading of TRU-only FCM fuel into a pin without significant quantities of uranium challenges the design from the standpoint of several key reactivity parameters, particularly void reactivity, and to some degree, the Doppler coefficient. These unit cells, while providing an indication of how a whole core of similar fuel would behave, also provide information of how individual pins of TRU-only FCM fuel would influence the reactivity behavior of a heterogeneous assembly. If these FCM fuel pins are included in a heterogeneous assembly with LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance of the TRU-only FCM fuel pins may be preserved. A configuration such as this would be similar to CONFU assemblies analyzed in previous studies. Analogous to the plutonium content limits imposed on MOX fuel, some amount of TRU-only FCM pins in an otherwise-uranium fuel assembly may give acceptable reactivity

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

    SciTech Connect

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

    1980-08-01

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

  19. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    SciTech Connect

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Schwalbach, P.; Sjoland, A.; Tobin, Stephen J.; Trellue, Holly; Vaccaro, S.

    2016-02-26

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  20. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    NASA Astrophysics Data System (ADS)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  1. Optimization of dry reforming of methane over Ni/YSZ anodes for solid oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Guerra, Cosimo; Lanzini, Andrea; Leone, Pierluigi; Santarelli, Massimo; Brandon, Nigel P.

    2014-01-01

    This work investigates the catalytic properties of Ni/YSZ anodes as electrodes of Solid Oxide Fuel Cells (SOFCs) to be operated under direct dry reforming of methane. The experimental test rig consists of a micro-reactor, where anode samples are characterized. The gas composition at the reactor outlet is monitored using a mass spectrometer. The kinetics of the reactions occurring over the anode is investigated by means of Isotherm reactions and Temperature-programmed reactions. The effect of the variation of temperature, gas residence time and inlet carbon dioxide-methane volumetric ratio is analyzed. At 800 °C, the best catalytic performance (in the carbon safe region) is obtained for 1.5 < carbon dioxide/methane ratio < 2, which is an interesting result for prospective direct biogas fueled SOFCs. Conversion is stable over a period of 70 h. Both for temperatures lower than 450 °C and for carbon dioxide-methane ratios lower than equi-molar at 800 °C, conversion is poor due to low activity of the anode toward dry reforming and cracking reactions, respectively. In other ranges, dry reforming and reverse water gas shift are the dominant reactions and the inlet feed reaches almost the equilibrium condition provided that a sufficient gas residence time is obtained.

  2. Remote sensing of fuel moisture content from canopy water indices and normalized dry matter index

    NASA Astrophysics Data System (ADS)

    Raymond Hunt, E.; Wang, Lingli; Qu, John J.; Hao, Xianjun

    2012-01-01

    Fuel moisture content (FMC), an important variable for predicting the occurrence and spread of wildfire, is the ratio of foliar water content and foliar dry matter content. One approach for the remote sensing of FMC has been to estimate the change in canopy water content over time by using a liquid-water spectral index. Recently, the normalized dry matter index (NDMI) was developed for the remote sensing of dry matter content using high-spectral-resolution data. The ratio of a spectral water index and a dry matter index corresponds to the ratio of foliar water and dry matter contents; therefore, we hypothesized that FMC may be remotely sensed with a spectral water index divided by NDMI. For leaf-scale simulations using the PROSPECT (leaf optical properties spectra) model, all water index/NDMI ratios were significantly related to FMC with a second-order polynomial regression. For canopy-scale simulations using the SAIL (scattering by arbitrarily inclined leaves) model, two water index/NDMI ratios, with numerators of the normalized difference infrared index (NDII) and the normalized difference water index (NDWI), predicted FMC with R2 values of 0.900 and 0.864, respectively. Leaves from three species were dried or stacked to vary FMC; measured NDII/NDMI was best related to FMC. Whereas the planned NASA mission Hyperspectral Infrared Imager (HyspIRI) will have high spectral resolution and very high signal-to-noise properties, the planned 19-day repeat frequency will not be sufficient for monitoring FMC with NDII/NDMI. Because increased fire frequency is expected with climatic change, operational assessment of FMC at large scales may require polar-orbiting environmental sensors with narrow bands to calculate NDMI.

  3. Dry spent fuel storage in Germany status in 1995 and prospects

    SciTech Connect

    Janberg, K. |; Malrnstroem, H.; Rittscher, D.; Willax, H.O.

    1995-12-31

    The German back-end policy until mid `94 was primarily based on reprocessing. Direct disposal was an acceptable alternative only when reprocessing was not available or economically not feasible. However, a law was passed in 1994 by Parliament which lifts these conditions applied to the choice of the final disposal route. For the THTR (Thorium High Temperature Reactor) fuel there was no reprocessing available and therefore the decommissioning of this reactor required the unloading of its fuel into dry storage casks. At the beginning of Nov `94 more than 260 CASTOR casks are already stored at the Ahaus site. The other storage facility at Gorleben was intended to be opened in July `94 with the CASTOR IIa, containing 4.5 t of HM. However, though the cask was loaded it is in early `95 waiting for its transport approval. The AVR-Reactor at the Juelich Research Center has been shut down and its fuel is also stored in casks. In early `95 around 50 are already loaded and transferred into the on-site storage facility. At the same time at the Greifswald site in former GDR a big storage facility is under construction. This facility has to receive all the wastes resulting from the decommissioning of the WWER 440 Voronesh-type reactors and the spent fuel also to be stored in casks.

  4. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    DOE PAGES

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

    2016-02-26

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

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

    SciTech Connect

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

    1984-04-01

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

  6. Select Generic Dry-Storage Pilot Plant Design for Safeguards and Security by Design (SSBD) per Used Fuel Campaign

    SciTech Connect

    Demuth, Scott Francis; Sprinkle, James K.

    2015-05-26

    As preparation to the year-end deliverable (Provide SSBD Best Practices for Generic Dry-Storage Pilot Scale Plant) for the Work Package (FT-15LA040501–Safeguards and Security by Design for Extended Dry Storage), the initial step was to select a generic dry-storage pilot plant design for SSBD. To be consistent with other DOE-NE Fuel Cycle Research and Development (FCR&D) activities, the Used Fuel Campaign was engaged for the selection of a design for this deliverable. For the work Package FT-15LA040501–“Safeguards and Security by Design for Extended Dry Storage”, SSBD will be initiated for the Generic Dry-Storage Pilot Scale Plant described by the layout of Reference 2. SSBD will consider aspects of the design that are impacted by domestic material control and accounting (MC&A), domestic security, and international safeguards.

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

    SciTech Connect

    Christensen, Max R; McKinnon, M. A.

    1999-12-01

    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.

  8. 40 CFR 80.141 - Interim detergent gasoline program.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 17 2013-07-01 2013-07-01 false Interim detergent gasoline program. 80... (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.141 Interim detergent gasoline... ultimate consumer; (ii) All additized post-refinery component (PRC); and (iii) All detergent additives...

  9. 40 CFR 80.141 - Interim detergent gasoline program.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 17 2014-07-01 2014-07-01 false Interim detergent gasoline program. 80... (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.141 Interim detergent gasoline... ultimate consumer; (ii) All additized post-refinery component (PRC); and (iii) All detergent additives...

  10. Used fuel extended storage security and safeguards by design roadmap

    SciTech Connect

    Durbin, Samuel G.; Lindgren, Eric Richard; Jones, Robert; Ketusky, Edward; England, Jeffrey; Scherer, Carolynn; Sprinkle, James; Miller, Michael.; Rauch, Eric; Scaglione, John; Dunn, T.

    2016-05-01

    In the United States, spent nuclear fuel (SNF) is safely and securely stored in spent fuel pools and dry storage casks. The available capacity in spent fuel pools across the nuclear fleet has nearly reached a steady state value. The excess SNF continues to be loaded in dry storage casks. Fuel is expected to remain in dry storage for periods beyond the initial dry cask certification period of 20 years. Recent licensing renewals have approved an additional 40 years. This report identifies the current requirements and evaluation techniques associated with the safeguards and security of SNF dry cask storage. A set of knowledge gaps is identified in the current approaches. Finally, this roadmap identifies known knowledge gaps and provides a research path to deliver the tools and models needed to close the gaps and allow the optimization of the security and safeguards approaches for an interim spent fuel facility over the lifetime of the storage site.

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

    SciTech Connect

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    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.

  12. Long-life high performance fuel cell program. Interim Report, 28 May 1981-31 October 1984

    SciTech Connect

    Martin, R.E.

    1985-02-01

    A multihundred kilowatt Regenerative Fuel Cell for use in a space station is envisioned. Three 0.508 sq ft (471.9 cm) active area multicell stacks were assembled and endurance tested. The long term performance stability of the platinum on carbon catalyst configuration suitability of the lightweight graphite electrolyte reservoir plate, the stability of the free standing butyl bonded potassium titanate matrix structure, and the long life potential of a hybrid polysulfone cell edge frame construction were demonstrated. A 18,000 hour demonstration test of multicell stack to a continuous cyclical load profile was conducted. A total of 12,000 cycles was completed, confirming the ability of the alkaline fuel cell to operate to a load profile simulating Regenerative Fuel Cell operation. An orbiter production hydrogen recirculation pump employed in support of the cyclical load profile test completed 13,000 hours of maintenance free operation. Laboratory endurance tests demonstrated the suitability of the butyl bonded potassium matrix, perforated nickel foil electrode substrates, and carbon ribbed substrate anode for use in the alkaline fuel cell. Corrosion testing of materials at 250 F (121.1 C) in 42% wgt. potassium identified ceria, zirconia, strontium titanate, strontium zirconate and lithium cobaltate as candidate matrix materials.

  13. Assessment of the integrity of spent fuel assemblies used in dry storage demonstrations at the Nevada Test Site

    SciTech Connect

    Johnson, A.B. Jr.; Dobbins, J.C.; Zaloudek, F.R.

    1987-07-01

    This report summarizes the histories of 17 Zircaloy-clad spent fuel assemblies used in dry storage tests and demonstrations at the Engine Maintenance and Disassembly (EMAD) and Climax facilities at the Nevada Test Site (NTS). The 18th assembly was shipped to the Battelle Columbus Laboratory (BCL) and remained there for extensive characterization and as a source of specimens for whole-rod and rod-segment dry storage tests. The report traces the history of the assemblies after discharge from the Turkey Point Unit 3 pressurized-water reactor (1975 and 1977) through shipment (first arrival at EMAD in December 1978), dry storage tests and demonstrations, and shipment by truck cask from EMAD to the Idaho National Engineering Laboratory (INEL) in May/June 1986. The principal objectives of this report are to assess and document the integrity of the fuel during the extensive dry storage activities at NTS and BCL, and to briefly summarize the dry storage technologies and procedures demonstrated in this program. The dry storage tests and demonstrations involved the following concepts and facilities: (1) surface drywells (EMAD); (2) deep drywells (425 m underground in the Climax granite formation); (3) concrete silo (EMAD); (4) air-cooled vault (EMAD); (5) electrically-heated module for fuel assembly thermal calibration and testing (EMAD/FAITM). 20 refs., 43 figs., 9 tabs.

  14. Processing of surrogate nuclear fuel pellets for better dimensional control with dry bag isostatic pressing

    NASA Astrophysics Data System (ADS)

    Hoggan, Rita E.; Zuck, Larry D.; Cannon, W. Roger; Lessing, Paul A.

    2016-12-01

    A study of improved methods of processing fuel pellets was undertaken using ceria and zirconia/yttria/alumina as surrogates. Through proper granulation, elimination of fines and vertical vibration (tapping) of the parts bag prior to dry bag isostatic pressing (DBIP), reproducibility of diameter profiles among multiple pellets of ceria was improved by almost an order of magnitude. Reproducibility of sintered pellets in these studies was sufficient to allow pellets to be introduced into the cladding with a gap between the pellet and cladding on the order of 50 μm to 100 μm but not a uniform gap with tolerance of ±12 μm as is currently required. Deviation from the mean diameter along the length of multiple pellets, and deviation from roundness, decreased after sintering. This is not generally observed with dry pressed pellets. Sintered shrinkage was uniform to ±0.05% and thus, as an alternative, pellets may be machined to tolerance before sintering, thus avoiding the waste associated with post-sinter grinding.

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

    SciTech Connect

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

    2013-09-01

    This report documents efforts to assess representative horizontal (Transuclear NUHOMS®) and vertical (Holtec HI-STORM) storage systems for the implementation of non-destructive examination (NDE) methods or techniques to manage atmospheric stress corrosion cracking (SCC) in canisters for dry storage of used nuclear fuel. The assessment is conducted by assessing accessibility and deployment, environmental compatibility, and applicability of NDE methods. A recommendation of this assessment is to focus on bulk ultrasonic and eddy current techniques for direct canister monitoring of atmospheric SCC. This assessment also highlights canister regions that may be most vulnerable to atmospheric SCC to guide the use of bulk ultrasonic and eddy current examinations. An assessment of accessibility also identifies canister regions that are easiest and more difficult to access through the ventilation paths of the concrete shielding modules. A conceivable sampling strategy for canister inspections is to sample only the easiest to access portions of vulnerable regions. There are aspects to performing an NDE inspection of dry canister storage system (DCSS) canisters for atmospheric SCC that have not been addressed in previous performance studies. These aspects provide the basis for recommendations of future efforts to determine the capability and performance of eddy current and bulk ultrasonic examinations for atmospheric SCC in DCSS canisters. Finally, other important areas of investigation are identified including the development of instrumented surveillance specimens to identify when conditions are conducive for atmospheric SCC, characterization of atmospheric SCC morphology, and an assessment of air flow patterns over canister surfaces and their influence on chloride deposition.

  16. Interaction of cosmic ray muons with spent nuclear fuel dry casks and determination of lower detection limit

    NASA Astrophysics Data System (ADS)

    Chatzidakis, S.; Choi, C. K.; Tsoukalas, L. H.

    2016-08-01

    The potential non-proliferation monitoring of spent nuclear fuel sealed in dry casks interacting continuously with the naturally generated cosmic ray muons is investigated. Treatments on the muon RMS scattering angle by Moliere, Rossi-Greisen, Highland and, Lynch-Dahl were analyzed and compared with simplified Monte Carlo simulations. The Lynch-Dahl expression has the lowest error and appears to be appropriate when performing conceptual calculations for high-Z, thick targets such as dry casks. The GEANT4 Monte Carlo code was used to simulate dry casks with various fuel loadings and scattering variance estimates for each case were obtained. The scattering variance estimation was shown to be unbiased and using Chebyshev's inequality, it was found that 106 muons will provide estimates of the scattering variances that are within 1% of the true value at a 99% confidence level. These estimates were used as reference values to calculate scattering distributions and evaluate the asymptotic behavior for small variations on fuel loading. It is shown that the scattering distributions between a fully loaded dry cask and one with a fuel assembly missing initially overlap significantly but their distance eventually increases with increasing number of muons. One missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the distributions which is the case of 100,000 muons. This indicates that the removal of a standard fuel assembly can be identified using muons providing that enough muons are collected. A Bayesian algorithm was developed to classify dry casks and provide a decision rule that minimizes the risk of making an incorrect decision. The algorithm performance was evaluated and the lower detection limit was determined.

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

    SciTech Connect

    Bryan, Charles R.; Enos, David

    2015-03-01

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

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

    SciTech Connect

    Santi, Peter Angelo; Browne, Michael C; Freeman, Corey R; Parker, Robert F; Williams, Richard B

    2010-01-01

    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

  19. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    SciTech Connect

    Mays, Gary T; Belles, Randy; Cetiner, Sacit M; Howard, Rob L; Liu, Cheng; Mueller, Don; Omitaomu, Olufemi A; Peterson, Steven K; Scaglione, John M

    2012-06-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  20. Analysis of liquid water formation in polymer electrolyte membrane (PEM) fuel cell flow fields with a dry cathode supply

    NASA Astrophysics Data System (ADS)

    Gößling, Sönke; Klages, Merle; Haußmann, Jan; Beckhaus, Peter; Messerschmidt, Matthias; Arlt, Tobias; Kardjilov, Nikolay; Manke, Ingo; Scholta, Joachim; Heinzel, Angelika

    2016-02-01

    PEM fuel cells can be operated within a wide range of different operating conditions. In this paper, the special case of operating a PEM fuel cell with a dry cathode supply and without external humidification of the cathode, is considered. A deeper understanding of the water management in the cells is essential for choosing the optimal operation strategy for a specific system. In this study a theoretical model is presented which aims to predict the location in the flow field at which liquid water forms at the cathode. It is validated with neutron images of a PEM fuel cell visualizing the locations at which liquid water forms in the fuel cell flow field channels. It is shown that the inclusion of the GDL diffusion resistance in the model is essential to describe the liquid water formation process inside the fuel cell. Good agreement of model predictions and measurement results has been achieved. While the model has been developed and validated especially for the operation with a dry cathode supply, the model is also applicable to fuel cells with a humidified cathode stream.

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

    SciTech Connect

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

    2005-07-27

    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.

  2. 75 FR 15420 - Interim Change to the Military Freight Traffic Unified Rules Publication (MFTURP) NO. 1

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-29

    ... Department of the Army Interim Change to the Military Freight Traffic Unified Rules Publication (MFTURP) NO... Command (SDDC) is providing notice that it will release an interim change to the MFTURP No. 1 on March 29, 2010. The interim change updates Section A, Paragraph N, Fuel Surcharge, in accordance with Section...

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

    SciTech Connect

    Krementz, Dan; Rose, David; Dunsmuir, Mike

    2014-02-06

    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

  4. Interim report

    SciTech Connect

    1985-06-01

    This Interim Report summarizes the research and development activities of the Superconducting Super Collider project carried out from the completion of the Reference Designs Study (May 1984) to June 1985. It was prepared by the SSC Central Design Group in draft form on the occasion of the DOE Annual Review, June 19--21, 1985. Now largely organized by CDG Divisions, the bulk of each chapter documents the progress and accomplishments to date, while the final section(s) describe plans for future work. Chapter 1, Introduction, provides a basic brief description of the SSC, its physics justification, its origins, and the R&D organization set up to carry out the work. Chapter 2 gives a summary of the main results of the R&D program, the tasks assigned to the four magnet R&D centers, and an overview of the future plans. The reader wishing a quick look at the SSC Phase I effort can skim Chapter 1 and read Chapter 2. Subsequent chapters discuss in more detail the activities on accelerator physics, accelerator systems, magnets and cryostats, injector, detector R&D, conventional facilities, and project planning and management. The magnet chapter (5) documents in text and photographs the impressive progress in successful construction of many model magnets, the development of cryostats with low heat leaks, and the improvement in current-carrying capacity of superconducting strand. Chapter 9 contains the budgets and schedules of the COG Divisions, the overall R&D program, including the laboratories, and also preliminary projections for construction. Appendices provide information on the various panels, task forces and workshops held by the CDG in FY 1985, a bibliography of COG and Laboratory reports on SSC and SSC-related work, and on private industrial involvement in the project.

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

    SciTech Connect

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

    1989-02-01

    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.

  6. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    SciTech Connect

    Bryan, Charles R.; Enos, David G.

    2015-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  7. Dry low NOx combustion system with pre-mixed direct-injection secondary fuel nozzle

    SciTech Connect

    Zuo, Baifang; Johnson, Thomas; Ziminsky, Willy; Khan, Abdul

    2013-12-17

    A combustion system includes a first combustion chamber and a second combustion chamber. The second combustion chamber is positioned downstream of the first combustion chamber. The combustion system also includes a pre-mixed, direct-injection secondary fuel nozzle. The pre-mixed, direct-injection secondary fuel nozzle extends through the first combustion chamber into the second combustion chamber.

  8. Economic analysis of fuel ethanol production from winter hulled barley by the EDGE (Enhanced Dry Grind Enzymatic) process.

    PubMed

    Nghiem, Nhuan P; Ramírez, Edna C; McAloon, Andrew J; Yee, Winnie; Johnston, David B; Hicks, Kevin B

    2011-06-01

    A process and cost model was developed for fuel ethanol production from winter barley based on the EDGE (Enhanced Dry Grind Enzymatic) process. In this process, in addition to β-glucanases, which are added to reduce the viscosity of the mash, β-glucosidase is also added to completely hydrolyze the oligomers obtained during the hydrolysis of β-glucans to glucose. The model allows determination of capital costs, operating costs, and ethanol production cost for a plant producing 40 million gallons of denatured fuel ethanol annually. A sensitivity study was also performed to examine the effects of β-glucosidase and barley costs on the final ethanol production cost. The results of this study clearly demonstrate the economic benefit of adding β-glucosidase. Lower ethanol production cost was obtained compared to that obtained without β-glucosidase addition in all cases except one where highest β-glucosidase cost allowance and lowest barley cost were used.

  9. Dry gas operation of proton exchange membrane fuel cells with parallel channels: Non-porous versus porous plates

    NASA Astrophysics Data System (ADS)

    Litster, Shawn; Santiago, Juan G.

    We present a study of proton exchange membrane (PEM) fuel cells with parallel channel flow fields for the cathode, dry inlet gases, and ambient pressure at the outlets. The study compares the performance of two designs: a standard, non-porous graphite cathode plate design and a porous hydrophilic carbon plate version. The experimental study of the non-porous plate is a control case and highlights the significant challenges of operation with dry gases and non-porous, parallel channel cathodes. These challenges include significant transients in power density and severe performance loss due to flooding and electrolyte dry-out. Our experimental study shows that the porous plate yields significant improvements in performance and robustness of operation. We hypothesize that the porous plate distributes water throughout the cell area by capillary action; including pumping water upstream to normally dry inlet regions. The porous plate reduces membrane resistance and air pressure drop. Further, IR-free polarization curves confirm operation free of flooding. With an air stoichiometric ratio of 1.3, we obtain a maximum power density of 0.40 W cm -2, which is 3.5 times greater than that achieved with the non-porous plate at the same operating condition.

  10. Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage

    SciTech Connect

    Tsoukalas, Lefteri H.

    2016-03-01

    This is the final report of the NEUP project “Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage”, DE-NE0000695. The project started on December 1, 2013 and this report covers the period December 1, 2013 through November 30, 2015. The project was successfully completed and this report provides an overview of the main achievements, results and findings throughout the duration of the project. Additional details can be found in the main body of this report and on the individual Quarterly Reports and associated Deliverables of the project, uploaded in PICS-NE.

  11. 40 CFR 80.155 - Interim detergent program controls and prohibitions.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 16 2011-07-01 2011-07-01 false Interim detergent program controls and... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.155 Interim detergent... wholesale purchaser-consumer, and no person shall detergent-additize gasoline, unless such gasoline...

  12. 40 CFR 80.155 - Interim detergent program controls and prohibitions.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 17 2012-07-01 2012-07-01 false Interim detergent program controls and... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.155 Interim detergent... wholesale purchaser-consumer, and no person shall detergent-additize gasoline, unless such gasoline...

  13. 40 CFR 80.155 - Interim detergent program controls and prohibitions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false Interim detergent program controls and... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.155 Interim detergent... wholesale purchaser-consumer, and no person shall detergent-additize gasoline, unless such gasoline...

  14. 40 CFR 80.155 - Interim detergent program controls and prohibitions.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 17 2014-07-01 2014-07-01 false Interim detergent program controls and... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.155 Interim detergent... wholesale purchaser-consumer, and no person shall detergent-additize gasoline, unless such gasoline...

  15. 40 CFR 80.155 - Interim detergent program controls and prohibitions.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 17 2013-07-01 2013-07-01 false Interim detergent program controls and... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.155 Interim detergent... wholesale purchaser-consumer, and no person shall detergent-additize gasoline, unless such gasoline...

  16. Effective thermal conductivity method for predicting spent nuclear fuel cladding temperatures in a dry fill gas

    SciTech Connect

    Bahney, Robert

    1997-12-19

    This paper summarizes the development of a reliable methodology for the prediction of peak spent nuclear fuel cladding temperature within the waste disposal package. The effective thermal conductivity method replaces other older methodologies.

  17. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  18. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  19. Spent nuclear fuel project cold vacuum drying facility vacuum and purge system design description

    SciTech Connect

    IRWIN, J.J.

    1998-11-30

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Vacuum and Purge System (VPS) . The SDD was developed in conjunction with HNF-SD-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-002, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the VPS equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  20. Spent nuclear fuel project cold vacuum drying facility process water conditioning system design description

    SciTech Connect

    IRWIN, J.J.

    1998-11-30

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Process Water Conditioning (PWC) System. The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), the HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the PWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  1. The Intentional Interim

    ERIC Educational Resources Information Center

    Nugent, Patricia A.

    2011-01-01

    The author spent years in central-office administration, most recently in an interim position. Some interim administrators simply see themselves as placeholders until the real deal is hired, giving the organization the opportunity to coast. There are others who see themselves as change agents and cannot wait to undo or redo what their predecessor…

  2. Hanford spent nuclear fuel cold vacuum drying proof of performance test procedure

    SciTech Connect

    McCracken, K.J.

    1998-06-10

    This document provides the test procedure for cold testing of the first article skids for the Cold Vacuum Drying (CVD) process at the Facility. The primary objective of this testing is to confirm design choices and provide data for the initial start-up parameters for the process. The current scope of testing in this document includes design verification, drying cycle determination equipment performance testing of the CVD process and MCC components, heat up and cool-down cycle determination, and thermal model validation.

  3. Drying, burning and emission characteristics of beehive charcoal briquettes: an alternative household fuel of Eastern Himalayan Region.

    PubMed

    Mandal, S; Kumar, Arvind; Singh, R K; Kundu, K

    2014-05-01

    Beehive charcoal briquettes were produced from powdered charcoal in which soil was added as binder. It was found to be an eco-friendly, clean and economic alternative source of household fuel for the people of Eastern Himalayan Region. Experiments were conducted to determine natural drying behaviour, normalised burn rate, temperature profile and emission of CO, CO2, UBHC (unburnt hydrocarbons) and NO(x) of beehive briquettes prepared from 60:40; 50:50 and 40:60 ratios of charcoal and soil. It was observed that under natural drying conditions (temperature, humidity) briquettes took 433 hr to reach equilibrium moisture content of 5.56-10.29%. Page's model was found suitable to describe the drying characteristics of all three combinations. Normalised burn rate varied between 0.377-0.706% of initial mass min⁻¹. Total burning time of briquette ranged between 133-143 min. The peak temperature attained by briquettes ranged from 437 °C to 572 °C. All the briquette combinations were found suitable for cooking and space heating. Emission of CO, CO2, UBHC, NO and NO2 ranged between 68.4-107.2, 922-1359, 20.9-50.8, 0.19-0.29 and 0.34-0.64 g kg⁻¹, respectively which were less than firewood.

  4. Preliminary safety evaluation for the spent nuclear fuel project`s cold vacuum drying system

    SciTech Connect

    Garvin, L.J., Westinghouse Hanford

    1996-07-01

    This preliminary safety evaluation (PSE) considers only the Cold Vacuum Drying System (CVDS) facility and its mission as it relates to the integrated process strategy (WHC 1995). The purpose of the PSE is to identify those CBDS design functions that may require safety- class and safety-significant accident prevention and mitigation features.

  5. Modeling of indirect carbon fuel cell systems with steam and dry gasification

    NASA Astrophysics Data System (ADS)

    Ong, Katherine M.; Ghoniem, Ahmed F.

    2016-05-01

    An indirect carbon fuel cell (ICFC) system that couples coal gasification to a solid oxide fuel cell (SOFC) is a promising candidate for high efficiency stationary power. This study couples an equilibrium gasifier model to a detailed 1D MEA model to study the theoretical performance of an ICFC system run on steam or carbon dioxide. Results show that the fuel cell in the ICFC system is capable of power densities greater than 1.0 W cm-2 with H2O recycle, and power densities ranging from 0.2 to 0.4 W cm-2 with CO2 recycle. This result indicates that the ICFC system performs better with steam than with CO2 gasification as a result of the faster electro-oxidation kinetics of H2 relative to CO. The ICFC system is then shown to reach higher current densities and efficiencies than a thermally decoupled gasifier + fuel cell (G + FC) system because it does not include combustion losses associated with autothermal gasification. 55-60% efficiency is predicted for the ICFC system coupled to a bottoming cycle, making this technology competitive with other state-of-the-art stationary power candidates.

  6. Membrane separation principle used for gas drying processes in fuel cells and life support systems

    NASA Astrophysics Data System (ADS)

    Nigsch, H. A.; Fleck, W. U.

    1991-07-01

    Different membrane separation principles as applied to fuel cell powerplants and ECLSS are described. A new separator type that enables smaller weight and geometries and requires less energy than conventional mechanical separator techniques for space applications is presented. Module optimization and investigations concerning ECLSS applications are discussed.

  7. Membrane separation principle used for gas drying processes in fuel cells and life support systems

    SciTech Connect

    Nigsch, H.A.; Fleck, W.U. )

    1991-07-01

    Different membrane separation principles as applied to fuel cell powerplants and ECLSS are described. A new separator type that enables smaller weight and geometries and requires less energy than conventional mechanical separator techniques for space applications is presented. Module optimization and investigations concerning ECLSS applications are discussed. 5 refs.

  8. Research development and demonstration of a fuel cell/battery powered bus system. Interim report, August 1, 1991--April 30, 1992

    SciTech Connect

    Romano, S.; Wimmer, R.

    1992-04-30

    This report describes the progress in the Georgetown University research, development and demonstration project of a fuel cell/battery powered bus system. The topics addressed in the report include vehicle design and application analysis, technology transfer activities, coordination and monitoring of system design and integration contractor, application of fuel cells to other vehicles, current problems, work planned, and manpower, cost and schedule reports.

  9. Cold vacuum drying facility: Phase 1 FMEA/FMECA session report

    SciTech Connect

    Pitkoff, C.C.

    1998-04-21

    The mission of the Spent Nuclear Fuel (SNF) Project is to remove the fuel currently located in the K-Basins 100 Area to provide safe handling and interim storage of the fuel. The spent nuclear fuel will be repackaged in multi-canister overpacks, partially dried in the Cold Vacuum Drying Facility (CVDF), and then transported to the Canister Storage Building (CSB) for further processing and interim storage. The CVDF, a subproject to the SNF Project, will be constructed in the 100K area. The CVDF will remove free water and vacuum dry the spent nuclear fuel, making it safer to transport and store at the CSB. At present, the CVDF is approximately 90% complete with definitive design. Part of the design process is to conduct Failure Modes, Effects, and Criticality Analysis (FMECA). A four-day FMECA session was conducted August 18 through 21, 1997. The purpose of the session was to analyze 16 subsystems and operating modes to determine consequences of normal, upset, emergency, and faulted conditions with respect to production and worker safety. During this process, acceptable and unacceptable risks, needed design or requirement changes, action items, issues/concerns, and enabling assumptions were identified and recorded. Additionally, a path forward consisting of recommended actions would be developed to resolve any unacceptable risks. The team consisted of project management, engineering, design authority, design agent, safety, operations, and startup personnel. The report summarizes potential problems with the designs, design requirements documentation, and other baseline documentation.

  10. Methods to recover value-added coproducts from dry grind processing of grains into fuel ethanol.

    PubMed

    Liu, Keshun; Barrows, Frederic T

    2013-07-31

    Three methods are described to fractionate condensed distillers solubles (CDS) into several new coproducts, including a protein-mineral fraction and a glycerol fraction by a chemical method; a protein fraction, an oil fraction and a glycerol-mineral fraction by a physical method; or a protein fraction, an oil fraction, a mineral fraction, and a glycerol fraction by a physicochemical method. Processing factors (ethanol concentration and centrifuge force) were also investigated. Results show that the three methods separated CDS into different fractions, with each fraction enriched with one or more of the five components (protein, oil, ash, glycerol and other carbohydrates) and thus having different targeted end uses. Furthermore, because glycerol, a hygroscopic substance, was mostly shifted to the glycerol or glycerol-mineral fraction, the other fractions had much faster moisture reduction rates than CDS upon drying in a forced air oven at 60 °C. Thus, these methods could effectively solve the dewatering problem of CDS, allowing elimination of the current industrial practice of blending distiller wet grains with CDS for drying together and production of distiller dried grains as a standalone coproduct in addition to a few new fractions.

  11. Dry additives-reduction catalysts for flue waste gases originating from the combustion of solid fuels

    SciTech Connect

    1995-12-31

    Hard coal is the basic energy generating raw material in Poland. In 1990, 60% of electricity and thermal energy was totally obtained from it. It means that 100 million tons of coal were burned. The second position is held by lignite - generating 38% of electricity and heat (67.3 million tons). It is to be underlined that coal combustion is particularly noxious to the environment. The coal composition appreciably influences the volume of pollution emitted in the air. The contents of incombustible mineral parts - ashes - oscillates from 2 to 30%; only 0.02 comes from plants that had once originated coal and cannot be separated in any way. All the rest, viz. the so-called external mineral substance enters the fuel while being won. The most indesirable hard coal ingredient is sulfur whose level depends on coal sorts and its origin. The worse the fuel quality, the more sulfur it contains. In the utilization process of this fuel, its combustible part is burnt: therefore, sulfur dioxide is produced. At the present coal consumption, the SO{sub 2} emission reaches the level of 3.2 million per year. The intensifies the pressure on working out new coal utilization technologies, improving old and developing of pollution limiting methods. Research is also directed towards such an adaptation of technologies in order that individual users may also make use thereof (household furnaces) as their share in the pollution emission is considerable.

  12. Dry halide method for separating the components of spent nuclear fuels

    DOEpatents

    Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

    1998-06-30

    The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

  13. Dry halide method for separating the components of spent nuclear fuels

    DOEpatents

    Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

    1998-01-01

    The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

  14. Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel

    SciTech Connect

    Gauld, I. C.; Ryman, J. C.

    2000-12-11

    This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance. The importance is investigated as a function of increasing burnup to assist in identifying the key changes in spent fuel characteristics between conventional- and extended-burnup regimes. Studies involving both pressurized water-reactor (PWR) fuel assemblies and boiling-water-reactor (BWR) assemblies are included. This study is seen to be a necessary first step in identifying the high-burnup spent fuel characteristics that may adversely affect the accuracy of current computational methods and data, assess the potential impact on previous guidance on isotopic source terms and decay-heat values, and thus help identify areas for methods and data improvement. Finally, several recommendations on the direction of possible future code validation efforts for high-burnup spent fuel predictions are presented.

  15. Numerical Modeling of Heat and Mass Transfer Processes in the Transfer of Spent Nuclear Fuel from "Wet" to "Dry" Cask Storage

    NASA Astrophysics Data System (ADS)

    Karyakin, Yu. E.; Pletnev, A. A.; Fedorovich, E. D.

    2017-01-01

    The paper describes in brief the heat and mass transfer processes in the transfer of spent nuclear fuel of the RBMK-100 reactor from "wet" to "dry" cask storage. The algorithms are described and the results are presented of the "through" calculation of the heat and mass transfer processes in ampoules and in a metal-concrete cask at various stages of spent nuclear fuel management.

  16. Toward a risk assessment of the spent fuel and high-level nuclear waste disposal system. Risk assessment requirements, literature review, methods evaluation: an interim report

    SciTech Connect

    Hamilton, L.D.; Hill, D.; Rowe, M.D.; Stern, E.

    1986-04-01

    This report provides background information for a risk assessment of the disposal system for spent nuclear fuel and high-level radioactive waste (HLW). It contains a literature review, a survey of the statutory requirements for risk assessment, and a preliminary evaluation of methods. The literature review outlines the state of knowledge of risk assessment and accident consequence analysis in the nuclear fuel cycle and its applicability to spent fuel and HLW disposal. The survey of statutory requirements determines the extent to which risk assessment may be needed in development of the waste-disposal system. The evaluation of methods reviews and evaluates merits and applicabilities of alternative methods for assessing risks and relates them to the problems of spent fuel and HLW disposal. 99 refs.

  17. Fuel Canister Stress Corrosion Cracking Susceptibility Experimental Results

    SciTech Connect

    Colleen Shelton-Davis

    2003-03-01

    The National Spent Nuclear Fuel Program is tasked with ensuring the U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) is acceptable for permanent disposal at a designated repository. From a repository acceptance criteria viewpoint and from a transportation viewpoint, of significant concern is the condition of the container at the time of shipment. Because the fuel will be in temporary storage for as much as 50 years, verification that no significant degradation has occurred to the canister is required to preclude repackaging all the fuel. Many canisters are also being removed from wet storage, vacuum dried (hot or cold), and then placed into dry storage. This process could have a detrimental effect on canister integrity. Research is currently underway to provide a technically sound assessment of the expected canister condition at the end of interim storage.

  18. CTR Fuel recovery system using regeneration of a molecular sieve drying bed

    DOEpatents

    Folkers, Charles L.

    1981-01-01

    A primary molecular sieve drying bed is regenerated by circulating a hot inert gas through the heated primary bed to desorb water held on the bed. The inert gas plus water vapor is then cooled and passed through an auxiliary molecular sieve bed which adsorbs the water originally desorbed from the primary bed. The main advantage of the regeneration technique is that the partial pressure of water can be reduced to the 10.sup.-9 atm. range. This is significant in certain CTR applications where tritiated water (T.sub.2 O, HTO) must be collected and kept at very low partial pressure.

  19. Nafion-porous cerium oxide nanotubes composite membrane for polymer electrolyte fuel cells operated under dry conditions

    NASA Astrophysics Data System (ADS)

    Ketpang, Kriangsak; Oh, Kwangjin; Lim, Sung-Chul; Shanmugam, Sangaraju

    2016-10-01

    A composite membrane operated in polymer electrolyte fuel cells (PEFCs) under low relative humidity (RH) is developed by incorporating cerium oxide nanotubes (CeNT) into a perfluorosulfonic acid (Nafion®) membrane. Porous CeNT is synthesized by direct heating a precursor impregnated polymer fibers at 500 °C under an air atmosphere. Compared to recast Nafion and commercial Nafion (NRE-212) membranes, the Nafion-CeNT composite membrane generates 1.1 times higher power density at 0.6 V, operated at 80 °C under 100% RH. Compared to Nafion-cerium oxide nanoparticles (Nafion-CeNP) membrane, the Nafion-CeNT provides 1.2 and 1.7 times higher PEFC performance at 0.6 V when operated at 80 °C under 100% and 18% RH, respectively. Additionally, the Nafion-CeNT composite membrane exhibits a good fuel cell operation under 18% RH at 80 °C. Specifically, the fluoride emission rate of Nafion-CeNT composite membrane is 20 times lower than that of the commercial NRE-212 membrane when operated under 18% RH at 80 °C for 96 h. The outstanding PEFC performance and durability operated under dry conditions is mainly attributed to the facile water diffusion capability as well as the effective hydroxyl radical scavenging property of the CeNT filler, resulting in significantly mitigating both the ohmic resistance and Nafion membrane degradation.

  20. Spent nuclear fuel project cold vacuum drying facility safety equipment list

    SciTech Connect

    IRWIN, J.J.

    1999-02-24

    This document provides the safety equipment list (SEL) for the Cold Vacuum Drying Facility (CVDF). The SEL was prepared in accordance with the procedure for safety structures, systems, and components (SSCs) in HNF-PRO-516, ''Safety Structures, Systems, and Components,'' Revision 0 and HNF-PRO-097, Engineering Design and Evaluation, Revision 0. The SEL was developed in conjunction with HNF-SO-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998). The SEL identifies the SSCs and their safety functions, the design basis accidents for which they are required to perform, the design criteria, codes and standards, and quality assurance requirements that are required for establishing the safety design basis of the SSCs. This SEL has been developed for the CVDF Phase 2 Safety Analysis Report (SAR) and shall be updated, expanded, and revised in accordance with future phases of the CVDF SAR until the CVDF final SAR is approved.

  1. High energy density proton exchange membrane fuel cell with dry reactant gases

    SciTech Connect

    Srinivasan, S.; Gamburzev, S.; Velev, O.A.

    1996-12-31

    Proton exchange membrane fuel cells (PEMFC) require careful control of humidity levels in the cell stack to achieve a high and stable level of performance. External humidification of the reactant gases, as in the state-of-the-art PEMFCs, increases the complexity, the weight, and the volume of the fuel cell power plant. A method for the operation of PEMFCs without external humidification (i.e., self-humidified PEMFCs) was first developed and tested by Dhar at BCS Technology. A project is underway in our Center to develop a PEMFC cell stack, which can work without external humidification and attain a performance level of a current density of 0.7 A/cm{sup 2} at a cell potential of 0.7 V, with hydrogen/air as reactants at 1 atm pressure. In this paper, the results of our efforts to design and develop a PEMFC stack requiring no external humidification will be presented. This paper focuses on determining the effects of type of electrodes, the methods of their preparation, as well as that of the membrane and electrode assembly (MEA), platinum loading and types of electrocatalyst on the performance of the PEMFC will be illustrated.

  2. Heat and Mass Transfer Modeling of Dry Gases in the Cathode of PEM Fuel Cells

    NASA Astrophysics Data System (ADS)

    Kermani, M. J.; Stockie, J. M.

    2004-02-01

    The transport of three gas species, O2, H2O and N2, through the cathode of a proton exchange membrane (PEM) fuel cell is studied numerically. The diffusion of the individual species is modeled via the Maxwell-Stefan equations, coupled with appropriate conservation equations. Two mechanisms are assumed for the internal energy sources in the system: a volumetric heat source due to the electrical current flowing through the cathode; and heat flow towards the cathode at the cathode-membrane interface due to the exothermic chemical reaction at this interface, in which water is generated. The governing equations of the unsteady fluid motion are written in fully conservative form, and consist of the following: (i) three equations for the mass conservation of the species; (ii) the momentum equation for the mixture, which is approximated using Darcy's Law for flow in porous media; and (iii) an energy equation, written in a form that has enthalpy as the dependent variable.

  3. Interim Joint Technical Assessment Report: Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards for Model Years 2017-2025

    EPA Pesticide Factsheets

    EPA and the NHTSA collaborated with CARB on this joint Technical Assessment Report to build on the success of the first phase of the National Program to regulate fuel economy and greenhouse gas (GHG) emissions from U.S. light-duty vehicles.

  4. Report of Ad Hoc Committee on Energy Efficiency in Transportation to the Interdepartmental Fuel and Energy Committee of the State of New York. Interim Report.

    ERIC Educational Resources Information Center

    New York State Interdepartmental Fuel and Energy Committee, Albany.

    After presenting the background of the availability of fuel for transportation and the increasing per capita energy consumption, the report examines the State's role in energy conservation. Five proposals are outlined: (1) a coordinated education program designed to increase public awareness of the current energy situation; (2) a pilot program of…

  5. West Valley vitrified HLW and spent-fuel on-site storage alternatives

    SciTech Connect

    Rothstein, H.; Swanson, J.; Kumar, S.

    1995-12-31

    Design layouts were developed for a West Valley Demonstration Project SSA with integrated interim storage of high-level radioactive waste canisters, spent fuel, and GTCC wastes from potential closure activities. Overall SSA cost estimates were prepared for the potential use of any of the NRC-licensed dry storage concepts. Using the costs for the concept closest to the average cost of all the concepts, comparisons were made to estimated costs for continued storage in the process building and FRS.

  6. Hanford K Basins spent nuclear fuels project update

    SciTech Connect

    Hudson, F.G.

    1997-10-17

    Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed.

  7. US PRACTICE FOR INTERIM WET STORAGE OF RRSNF

    SciTech Connect

    Vinson, D.

    2010-08-05

    Aluminum research reactor spent nuclear fuel is currently being stored or is anticipated to be returned to the United States and stored at Department of Energy storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper summarizes the current practices to provide for continued safe interim wet storage in the U.S. Aluminum fuel stored in poor quality water is subject to aggressive corrosion attack and therefore water chemistry control systems are essential to maintain water quality. Fuel with minor breaches are safely stored directly in the basin. Fuel pieces and heavily damaged fuel is safely stored in isolation canisters.

  8. Definition of chemical and electrochemical properties of a fuel cell electrolyte. Interim technical report, 24 July 1978-24 December 1979

    SciTech Connect

    Ahmad, J.; Foley, R.T.

    1980-01-01

    The present research is oriented toward the task of developing an improved electrolyte for the direct hydrocarbon-air fuel cell. The electrochemical behavior of methanesulfonic acid, ethanesulfonic acid, and sulfoacetic acid as fuel cell electrolytes was studied in a half cell at various temperatures. The rate of electro-oxidation of hydrogen at 115 degrees was very high in methanesulfonic acid and sulfoacetic acids. The rate of the electro-oxidation of propane in methanesulfonic acid at 80/sup 0/C and 115/sup 0/C was low. Further, there is evidence for adsorption of these acids on the platinum electrode. Sulfoacetic acid with H/sup 2/ has supported about two times higher current density than trifluoromethanesulfonic acid monohydrate, but, attempts to purify the compound were unsuccessful. It was concluded that anhydrous sulfonic acids are not good electrolytes; water solutions are required. Sulfonic acids containing unprotected C-H bonds are adsorbed on platinum and probably decompose during electrolysis. A completely substituted sulfonic acid would be the preferred electrolyte.

  9. Online Education Interim Report

    ERIC Educational Resources Information Center

    Protopsaltis, Spiros, Ed.

    2007-01-01

    This interim report produced by the Colorado State Board of Education Online Education Task Force examines key issues related to online education. Task force members agree that: (1) online education has become a viable element of Colorado's public education system; (2) the role of technology in educating our children will continue to grow; and (3)…

  10. Interim Budget Plan.

    ERIC Educational Resources Information Center

    Office of Student Financial Assistance (ED), Washington, DC.

    This report provides the interim budget plan of the U.S. Department of Education's Office of Student Financial Assistance (OSFA) for fiscal year 2000. It reviews factors influencing OSFA's budget request, including legislative requirements, recent accomplishments, the need to maintain both the Direct Loan and Federal Family Education Loan…

  11. Used Nuclear Fuel: From Liability to Benefit

    NASA Astrophysics Data System (ADS)

    Orbach, Raymond L.

    2011-03-01

    Nuclear power has proven safe and reliable, with operating efficiencies in the U.S. exceeding 90%. It provides a carbon-free source of electricity (with about a 10% penalty arising from CO2 released from construction and the fuel cycle). However, used fuel from nuclear reactors is highly toxic and presents a challenge for permanent disposal -- both from technical and policy perspectives. The half-life of the ``bad actors'' is relatively short (of the order of decades) while the very long lived isotopes are relatively benign. At present, spent fuel is stored on-site in cooling ponds. Once the used fuel pools are full, the fuel is moved to dry cask storage on-site. Though the local storage is capable of handling used fuel safely and securely for many decades, the law requires DOE to assume responsibility for the used fuel and remove it from reactor sites. The nuclear industry pays a tithe to support sequestration of used fuel (but not research). However, there is currently no national policy in place to deal with the permanent disposal of nuclear fuel. This administration is opposed to underground storage at Yucca Mountain. There is no national policy for interim storage---removal of spent fuel from reactor sites and storage at a central location. And there is no national policy for liberating the energy contained in used fuel through recycling (separating out the fissionable components for subsequent use as nuclear fuel). A ``Blue Ribbon Commission'' has been formed to consider alternatives, but will not report until 2012. This paper will examine alternatives for used fuel disposition, their drawbacks (e.g. proliferation issues arising from recycling), and their benefits. For recycle options to emerge as a viable technology, research is required to develop cost effective methods for treating used nuclear fuel, with attention to policy as well as technical issues.

  12. Hanford spent nuclear fuel project update

    SciTech Connect

    Williams, N.H.

    1997-08-19

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

  13. CMM Interim Check (U)

    SciTech Connect

    Montano, Joshua Daniel

    2015-03-23

    Coordinate Measuring Machines (CMM) are widely used in industry, throughout the Nuclear Weapons Complex and at Los Alamos National Laboratory (LANL) to verify part conformance to design definition. Calibration cycles for CMMs at LANL are predominantly one year in length. Unfortunately, several nonconformance reports have been generated to document the discovery of a certified machine found out of tolerance during a calibration closeout. In an effort to reduce risk to product quality two solutions were proposed – shorten the calibration cycle which could be costly, or perform an interim check to monitor the machine’s performance between cycles. The CMM interim check discussed makes use of Renishaw’s Machine Checking Gauge. This off-the-shelf product simulates a large sphere within a CMM’s measurement volume and allows for error estimation. Data was gathered, analyzed, and simulated from seven machines in seventeen different configurations to create statistical process control run charts for on-the-floor monitoring.

  14. Determination of the boundary of carbon formation for dry reforming of methane in a solid oxide fuel cell

    NASA Astrophysics Data System (ADS)

    Assabumrungrat, S.; Laosiripojana, N.; Piroonlerkgul, P.

    The boundary of carbon formation for the dry reforming of methane in direct internal reforming solid oxide fuel cells (DIR-SOFCs) with different types of electrolyte (i.e., an oxygen ion-conducting electrolyte (SOFC-O 2-) and a proton-conducting electrolyte (SOFC-H +)) was determined by employing detailed thermodynamic analysis. It was found that the required CO 2/CH 4 ratio decreased with increasing temperature. The type of electrolyte influenced the boundary of carbon formation because it determined the location of water formed by the electrochemical reaction. The extent of the electrochemical reaction also played an important role in the boundary of carbon formation. For SOFC-O 2-, the required CO 2/CH 4 ratio decreased with the increasing extent of the electrochemical reaction due to the presence of electrochemical water in the anode chamber. Although for SOFC-H + the required CO 2/CH 4 ratio increased with the increasing extent of the electrochemical reaction at high operating temperature (T > 1000 K) following the trend previously reported for the case of steam reforming of methane with addition of water as a carbon suppresser, an unusual opposite trend was observed at lower operating temperature. The study also considered the use of water or air as an alternative carbon suppresser for the system. The required H 2O/CH 4 ratio and air/CH 4 ratio were determined for various inlet CO 2/CH 4 ratios. Even air is a less attractive choice compared to water due to the higher required air/CH 4 ratio than the H 2O/CH 4 ratio; however, the integration of exothermic oxidation and the endothermic reforming reactions may make the use of air attractive. Water was found to be more effective than carbon dioxide in suppressing the carbon formation at low temperatures but their effect was comparable at high temperatures. Although the results from the study were based on calculations of the SOFCs with different electrolytes, they are also useful for selecting suitable feed

  15. Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

    PubMed

    Rezaeian, M; Kamali, J

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B4C) were investigated, and the minimum required receptacle pitch of the basket was determined.

  16. Immobilized High Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report 2nd Generation Implementing Architecture

    SciTech Connect

    CALMUS, R.B.

    2000-09-14

    Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document.

  17. Feasibility study for Zaporozhye Nuclear Power Plant spent fuel dry storage facility in Ukraine. Export trade information

    SciTech Connect

    1995-12-01

    This document reports the results of a Feasibility Study sponsored by a TDA grant to Zaporozhye Nuclear Power Plant (ZNPP) in Ukraine to study the construction of storage facilities for spent nuclear fuel. It provides pertinent information to U.S. companies interested in marketing spent fuel storage technology and related business to countries of the former Soviet Union or Eastern Europe.

  18. Interim Assessments: A User's Guide

    ERIC Educational Resources Information Center

    Marshall, Kim

    2008-01-01

    Interim assessments are hot in American schools. Also called benchmark or periodic tests, these assessments are given every four to nine weeks to check on students' progress. Although interim assessments are an important tool for school improvement, they are easy to misuse. In the author's work coaching principals in a number of districts, he has…

  19. Guidance: Interim Municipal Settlement Policy

    EPA Pesticide Factsheets

    Interim guidance and fact sheets regarding settlements involving municipalities or municipal waste under Section 122 CERCLA as amended by SARA. Interim policy sets forth the criteria by which EPA generally determines whether to exercise enforcement discretion to pursue MSW generators and transporters as PRPs.

  20. Evaluation of advanced combustion concepts for dry NO sub x suppression with coal-derived, gaseous fuels

    NASA Technical Reports Server (NTRS)

    Beebe, K. W.; Symonds, R. A.; Notardonato, J. J.

    1982-01-01

    The emissions performance of a rich lean combustor (developed for liquid fuels) was determined for combustion of simulated coal gases ranging in heating value from 167 to 244 Btu/scf (7.0 to 10.3 MJ/NCM). The 244 Btu/scf gas is typical of the product gas from an oxygen blown gasifier, while the 167 Btu/scf gas is similar to that from an air blown gasifier. NOx performance of the rich lean combustor did not meet program goals with the 244 Btu/scf gas because of high thermal NOx, similar to levels expected from conventional lean burning combustors. The NOx emissions are attributed to inadequate fuel air mixing in the rich stage resulting from the design of the large central fuel nozzle delivering 71% of the total gas flow. NOx yield from ammonia injected into the fuel gas decreased rapidly with increasing ammonia level, and is projected to be less than 10% at NH3 levels of 0.5% or higher. NOx generation from NH3 is significant at ammonia concentrations significantly less than 0.5%. These levels may occur depending on fuel gas cleanup system design. CO emissions, combustion efficiency, smoke and other operational performance parameters were satisfactory. A test was completed with a catalytic combustor concept with petroleum distillate fuel. Reactor stage NOx emissions were low (1.4g NOx/kg fuel). CO emissions and combustion efficiency were satisfactory. Airflow split instabilities occurred which eventually led to test termination.

  1. Interim storage study report

    SciTech Connect

    Rawlins, J.K.

    1998-02-01

    High-level radioactive waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) in the form of calcine and liquid and liquid sodium-bearing waste (SBW) will be processed to provide a stable waste form and prepare the waste to be transported to a permanent repository. Because a permanent repository will not be available when the waste is processed, the waste must be stored at ICPP in an Interim Storage Facility (ISF). This report documents consideration of an ISF for each of the waste processing options under consideration.

  2. Spent Fuel Background Report Volume I

    SciTech Connect

    Abbott, D.

    1994-03-01

    This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial reactors used in research activities at DOE sites. Naval fuels are those developed and used for nuclear-powered naval vessels and for related research and development. Given the recent DOE decision to curtail reprocessing, the topic of main concern in the management of spent fuel is its storage. Of the DOE sites that have spent nuclear fuel, the vast majority is located at three sites-Hanford, INEL, and Savannah River. Other sites with spent fuel include Oak Ridge, West Valley, Brookhaven, Argonne, Los Alamos, and Sandia. B&W NESI Lynchburg Technology Center and General Atomics are commercial facilities with DOE fuel. DOE may also receive fuel from foreign research reactors, university reactors, and other commercial and government research reactors. Most DOE spent fuel is stored in water-filled pools at the reactor facilities. Currently an engineering study is being performed to determine the feasibility of using dry storage for DOE-owned spent fuel currently stored at various facilities. Delays in opening the deep geologic

  3. 76 FR 4369 - Interim Deputation Agreements; Interim BIA Adult Detention Facility Guidelines

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-25

    ... Bureau of Indian Affairs Interim Deputation Agreements; Interim BIA Adult Detention Facility Guidelines... publication of the Interim BIA Adult Detention Facility Guidelines and the Interim Model Deputation Agreements... of 2010. Three Interim Model Deputation Agreements will be used: one agreement for tribes in...

  4. Multidimensional simulations of hydrides during fuel rod lifecycle

    NASA Astrophysics Data System (ADS)

    Stafford, D. S.

    2015-11-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

  5. 49 CFR 1182.7 - Interim approval.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 8 2010-10-01 2010-10-01 false Interim approval. 1182.7 Section 1182.7....7 Interim approval. (a) A party may request interim approval of the operation of the properties... additional filing fee is required, whether the request for interim approval is included in the application...

  6. 24 CFR 35.820 - Interim controls.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Interim controls. 35.820 Section 35...-Possession Multifamily Property § 35.820 Interim controls. HUD shall conduct interim controls in accordance... accordance with § 35.815. Interim controls are considered completed when clearance is achieved in...

  7. Ni-(Ce0.8-xTix)Sm0.2O2-δ anode for low temperature solid oxide fuel cells running on dry methane fuel

    NASA Astrophysics Data System (ADS)

    Han, Bing; Zhao, Kai; Hou, Xiaoxue; Kim, Dong-Jin; Kim, Bok-Hee; Ha, Su; Norton, M. Grant; Xu, Qing; Ahn, Byung-Guk

    2017-01-01

    A titanium-doped Ce0.8Sm0.2O1.9 composite is developed as an anode component of low temperature solid oxide fuel cells running on methane fuel. Crystallographic parameters of (Ce0.8-xTix)Sm0.2O2-δ (0.00 < x < 0.10) are investigated with respect to the amount of titanium. The composites show a single cubic phase with the titanium amount being in the range of 0.00-0.07, while the lattice parameters decrease with increasing titanium content. The (Ce0.8-xTix)Sm0.2O2-δ composites are applied to an anode-supported single cell consisting of Ni-(Ce0.8-xTix)Sm0.2O2-δ anode/Ce0.8Sm0.2O1.9 electrolyte/La0.6Sr0.4Co0.2Fe0.8O3-δ cathode. Catalytic properties of Ni-(Ce0.8-xTix)Sm0.2O2-δ are inspected with the electrochemical performance and performance stability of the cells in dry methane fuel. The cell with Ni-(Ce0.73Ti0.07)Sm0.2O2-δ (x = 0.07) anode displays a low polarization resistance and an optimum maximum power density (679 mW cm-2 at 600 °C). A performance stability investigation indicates that the cell exhibits a fairly low degradation rate of 3 mV h-1 during a 31 h operation in dry methane. These findings suggest the application potential of the titanium doped Ce0.8Sm0.2O1.9 for the anode of solid oxide fuel cells.

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

    SciTech Connect

    Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y.

    2012-07-06

    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

  9. Spent nuclear fuel project cold vacuum drying facility tempered water and tempered water cooling system design description

    SciTech Connect

    IRWIN, J.J.

    1998-11-30

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Tempered Water (TW) and Tempered Water Cooling (TWC) System . The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the TW and TWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SOD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  10. Evaluation of hoop creep behaviors in long-term dry storage condition of pre-hydrided and high burn-up nuclear fuel cladding

    SciTech Connect

    Kim, Sun-Ki; Bang, J.G.; Kim, D.H.; Yang, Y.S.

    2007-07-01

    Related to the degradation of the mechanical properties of Zr-based nuclear fuel cladding tubes under long term dry storage condition, the mechanical tests which can simulate the degradation of the mechanical properties properly are needed. Especially, the degradation of the mechanical properties by creep mechanism seems to be dominant under long term dry storage condition. Accordingly, in this paper, ring creep tests were performed in order to evaluate the creep behaviors of high burn-up fuel cladding under a hoop loading condition in a hot cell. The tests are performed with Zircaloy-4 fuel cladding whose burn-up is approximately {approx}60,000 MWd/tU in the temperature range from 350 deg. to 550 deg.. The tests are also performed with pre-hydrided Zircaloy-4 and ZIRLO up to 1,000 ppm. First of all, the hoop loading grip for the ring creep test was designed in order that a constant curvature of the specimen was maintained during the creep deformation, and the graphite lubricant was used to minimize the friction between the outer surface of the die insert and the inner surface of the ring specimen. The specimen for the ring creep test was designed to limit the deformation within the gauge section and to maximize the uniformity of the strain distribution. It was confirmed that the mechanical properties under a hoop loading condition can be correctly evaluated by using this test technique. In this paper, secondary creep rate with increasing hydrogen content are drawn, and then kinetic data such as pre-exponential factor and activation energy for creep process are also drawn. In addition, creep life are predicted by obtaining LMP (Larson-Miller parameter) correlation in the function of hydrogen content and applied stress to yield stress ratio. (authors)

  11. AGR-1 Data Qualification Interim Report

    SciTech Connect

    Machael Abbott

    2009-08-01

    Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010.

  12. Method of preparing nuclear wastes for tansportation and interim storage

    DOEpatents

    Bandyopadhyay, Gautam; Galvin, Thomas M.

    1984-01-01

    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.

  13. Interim dentures and treatment dentures.

    PubMed

    Smith, D E

    1984-04-01

    Improvement in the interim denture procedure in the past decade has been one of the significant advancements in prosthodontic practice. The interim denture approach is only slightly more time-consuming and expensive than the conventional immediate denture approach, yet it has many advantages. Among those advantages are the following: (1) allows rapid results; (2) results in a higher quality definitive denture; (3) allows the surgical treatment to be performed during one appointment; (4) permits duplication of the natural tooth position; and (5) provides the patient with a spare denture after the definitive denture is completed. An interim denture technique was described that utilized a flexible layered silicone mold to form the replaced teeth. The interim denture procedure is flexible and lends itself to many variations in technique to meet unusual clinical situations. An interim removable partial denture technique was described that involves block-out of undesirable undercuts and duplication of the master cast for fabrication of the partial denture. This technique results in an interim partial denture that can be placed with little or no adjustment and that will provide better service for those who require it. Three simple procedures for fabricating treatment dentures were described and the indications for each were discussed. Although treatment dentures are not often used, they are essential for the dentist who is treating difficult patients who require complete dentures.

  14. An Evaluation of the Functionality of Advanced Fuel Research Prototype Dry Pyrolyzer for Destruction of Solid Wastes

    NASA Technical Reports Server (NTRS)

    Fisher, John; Wignarajah, K.; Howard, Kevin; Serio, Mike; Kroo, Eric

    2004-01-01

    The prototype dry pyrolyser delivered to Ames Research Center is the end-product of a Phase I1 Small Business Initiative Research (SBIR) project. Some of the major advantages of pyrolysis for processing solid wastes are that it can process solid wastes, it permits elemental recycling while conserving oxygen use, and it can function as a pretreatment for combustion processes. One of the disadvantages of pyrolysis is the formation of tars. By controlling the rate of heating, tar formation can be minimized. This paper presents data on the pyrolysis of various space station wastes. The performance of the pyrolyser is also discussed and appropriate modifications suggested to improve the performance of the dry pyrolyzer.

  15. Catalytic modification of Ni-Sm-doped ceria anodes with copper for direct utilization of dry methane in low-temperature solid oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Wang, Zhicheng; Weng, Wenjian; Cheng, Kui; Du, Piyi; Shen, Ge; Han, Gaorong

    2008-05-01

    A Cu/Ni/Sm-doped ceria (SDC) anode has been designed for direct utilization of dry methane in low-temperature anode-supported solid oxide fuel cells. The anode is prepared by the impregnation method, whereby a small amount of Cu is incorporated into the previously prepared Ni/SDC porous matrix. After reduction, Cu nanoparticles adhere to and are uniformly distributed on the surface of the Ni/SDC matrix. For the resulting Cu/Ni/SDC anode-supported cell, maximum power density of 317 mW cm-2 is achieved at 600 °C. The power density shows only ∼2% loss after 12-h operation. The results demonstrate that the Cu/Ni/SDC anode effectively suppresses carbon deposition by decreasing the Ni surface area available and the level of carbon monoxide disproportionation. This combination of effects results in very low-power density loss over the operating time.

  16. Investigation of sulfur interactions on a conventional nickel-based solid oxide fuel cell anode during methane steam and dry reforming

    NASA Astrophysics Data System (ADS)

    Jablonski, Whitney S.

    Solid oxide fuel cells (SOFC) are an attractive energy source because they do not have undesirable emissions, are scalable, and are feedstock flexible, which means they can operate using a variety of fuel mixtures containing H2 and hydrocarbons. In terms of fuel flexibility, most potential fuel sources contain sulfur species, which severely poison the nickel-based anode. The main objective of this thesis is to systematically evaluate sulfur interactions on a conventional Ni/YSZ anode and compare sulfur poisoning during methane steam and dry reforming (SMR and DMR) to a conventional catalyst (Sud Chemie, Ni/K2O-CaAl2O4). Reforming experiments (SMR and DMR) were carried out in a packed bed reactor (PBR), and it was demonstrated that Ni/YSZ is much more sensitive to sulfur poisoning than Ni/K2O-CaAl2O4 as evidenced by the decline in activity to zero in under an hour for both SMR and DMR. Adsorption and desorption of H2S and SO2 on both catalysts was evaluated, and despite the low amount of accessible nickel on Ni/YSZ (14 times lower than Ni/K2O-CaAl2O4), it adsorbs 20 times more H2S and 50 times more SO2 than Ni/K 2O-CaAl2O4. A one-dimensional, steady state PBR model (DetchemPBED) was used to evaluate SMR and DMR under poisoning conditions using the Deutschmann mechanism and a recently published sulfur sub-mechanism. To fit the observed deactivation in the presence of 1 ppm H2S, the adsorption/desorption equilibrium constant was increased by a factor 16,000 for Ni/YSZ and 96 for Ni/K2O-CaAl2O4. A tubular SAE reactor was designed and fabricated for evaluating DMR in a reactor that mimics an SOFC. Evidence of hydrogen diffusion through a supposedly impermeable layer indicated that the tubular SAE reactor has a major flaw in which gases diffuse to unintended parts of the tube. It was also found to be extremely susceptible to coking which leads to cell failure even in operating regions that mimic real biogas. These problems made it impossible to validate the tubular SAE

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

    SciTech Connect

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

    2013-04-01

    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.

  18. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  19. Analysis of Ignition Testing on K-West Basin Fuel

    SciTech Connect

    J. Abrefah; F.H. Huang; W.M. Gerry; W.J. Gray; S.C. Marschman; T.A. Thornton

    1999-08-10

    Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basin into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994).

  20. Dry Mouth

    MedlinePlus

    ... of this page please turn Javascript on. Dry Mouth What Is Dry Mouth? Dry mouth is the feeling that there is ... when a person has dry mouth. How Dry Mouth Feels Dry mouth can be uncomfortable. Some people ...

  1. FedEx Gasoline Hybrid Electric Delivery Truck Evaluation: 6-Month Interim Report

    SciTech Connect

    Barnitt, R.

    2010-05-01

    This interim report presents partial (six months) results for a technology evaluation of gasoline hybrid electric parcel delivery trucks operated by FedEx in and around Los Angeles, CA. A 12 month in-use technology evaluation comparing in-use fuel economy and maintenance costs of GHEVs and comparative diesel parcel delivery trucks was started in April 2009. Comparison data was collected and analyzed for in-use fuel economy and fuel costs, maintenance costs, total operating costs, and vehicle uptime. In addition, this interim report presents results of parcel delivery drive cycle collection and analysis activities as well as emissions and fuel economy results of chassis dynamometer testing of a gHEV and a comparative diesel truck at the National Renewable Energy Laboratory's (NREL) ReFUEL laboratory. A final report will be issued when 12 months of in-use data have been collected and analyzed.

  2. 7 CFR 15a.71 - Interim procedures.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 7 Agriculture 1 2013-01-01 2013-01-01 false Interim procedures. 15a.71 Section 15a.71 Agriculture Office of the Secretary of Agriculture EDUCATION PROGRAMS OR ACTIVITIES RECEIVING OR BENEFITTING FROM FEDERAL FINANCIAL ASSISTANCE Procedures (Interim) § 15a.71 Interim procedures. For the purposes...

  3. 7 CFR 15a.71 - Interim procedures.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 7 Agriculture 1 2012-01-01 2012-01-01 false Interim procedures. 15a.71 Section 15a.71 Agriculture Office of the Secretary of Agriculture EDUCATION PROGRAMS OR ACTIVITIES RECEIVING OR BENEFITTING FROM FEDERAL FINANCIAL ASSISTANCE Procedures (Interim) § 15a.71 Interim procedures. For the purposes...

  4. 7 CFR 15a.71 - Interim procedures.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 7 Agriculture 1 2014-01-01 2014-01-01 false Interim procedures. 15a.71 Section 15a.71 Agriculture Office of the Secretary of Agriculture EDUCATION PROGRAMS OR ACTIVITIES RECEIVING OR BENEFITTING FROM FEDERAL FINANCIAL ASSISTANCE Procedures (Interim) § 15a.71 Interim procedures. For the purposes...

  5. 7 CFR 15a.71 - Interim procedures.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 1 2011-01-01 2011-01-01 false Interim procedures. 15a.71 Section 15a.71 Agriculture Office of the Secretary of Agriculture EDUCATION PROGRAMS OR ACTIVITIES RECEIVING OR BENEFITTING FROM FEDERAL FINANCIAL ASSISTANCE Procedures (Interim) § 15a.71 Interim procedures. For the purposes...

  6. 7 CFR 1735.75 - Interim financing.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 11 2010-01-01 2010-01-01 false Interim financing. 1735.75 Section 1735.75... Involving Loan Funds § 1735.75 Interim financing. (a) A borrower may submit a written request for RUS approval of interim financing if it is necessary to close an acquisition before the loan to finance...

  7. 20 CFR 801.202 - Interim appointments.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 20 Employees' Benefits 3 2010-04-01 2010-04-01 false Interim appointments. 801.202 Section 801.202... Members of the Board § 801.202 Interim appointments. (a) Acting Chairman. In the event that the Chairman... such time as the Secretary of Labor designates an Acting Chairman. (b) Interim members. In the...

  8. 24 CFR 7.44 - Interim relief.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Interim relief. 7.44 Section 7.44... § 7.44 Interim relief. (a) When the Department appeals and the case involves removal, separation, or... outcome of the Department appeal. The employee may decline the offer of interim relief. (b) Service...

  9. 7 CFR 15a.71 - Interim procedures.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 1 2010-01-01 2010-01-01 false Interim procedures. 15a.71 Section 15a.71 Agriculture Office of the Secretary of Agriculture EDUCATION PROGRAMS OR ACTIVITIES RECEIVING OR BENEFITTING FROM FEDERAL FINANCIAL ASSISTANCE Procedures (Interim) § 15a.71 Interim procedures. For the purposes...

  10. 7 CFR 1738.21 - Interim financing.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 11 2010-01-01 2010-01-01 false Interim financing. 1738.21 Section 1738.21... Interim financing. (a) Upon notification by RUS that an applicant's application is considered complete, the applicant may enter into an interim financing agreement with a lender other than RUS or use...

  11. 15 CFR 904.322 - Interim action.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Interim action. 904.322 Section 904... Sanctions and Denials Permit Sanction for Violations § 904.322 Interim action. (a) To protect marine...) The Judge will order interim action under paragraph (a) of this section, only after finding that...

  12. 15 CFR 908.5 - Interim reports.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Interim reports. 908.5 Section 908.5... SUBMITTING REPORTS ON WEATHER MODIFICATION ACTIVITIES § 908.5 Interim reports. (a) Any person engaged in a... Administrator, not later than 45 days thereafter, an interim report setting forth as of such date...

  13. 19 CFR 207.106 - Interim measures.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 19 Customs Duties 3 2010-04-01 2010-04-01 false Interim measures. 207.106 Section 207.106 Customs... and Committee Proceedings § 207.106 Interim measures. (a) At any time after proceedings are initiated... that would otherwise be kept confidential, or to take other appropriate interim measures. (b)...

  14. 12 CFR 268.505 - Interim relief.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 12 Banks and Banking 3 2010-01-01 2010-01-01 false Interim relief. 268.505 Section 268.505 Banks... REGARDING EQUAL OPPORTUNITY Remedies and Enforcement § 268.505 Interim relief. (a)(1) When the Board appeals... offer of interim relief. (2) Service under the temporary or conditional restoration provisions...

  15. 42 CFR 417.574 - Interim settlement.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 42 Public Health 3 2010-10-01 2010-10-01 false Interim settlement. 417.574 Section 417.574 Public... PLANS Medicare Payment: Cost Basis § 417.574 Interim settlement. (a) Determination. Within 30 days following the receipt of the HMO's or CMP's final interim cost and enrollment reports, CMS will make...

  16. 19 CFR 354.8 - Interim sanctions.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 19 Customs Duties 3 2010-04-01 2010-04-01 false Interim sanctions. 354.8 Section 354.8 Customs... VIOLATION OF AN ANTIDUMPING OR COUNTERVAILING DUTY ADMINISTRATIVE PROTECTIVE ORDER § 354.8 Interim sanctions... a final decision is rendered, that interim sanctions are necessary to protect the interests of...

  17. 19 CFR 356.18 - Interim sanctions.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 19 Customs Duties 3 2010-04-01 2010-04-01 false Interim sanctions. 356.18 Section 356.18 Customs... Undertaking § 356.18 Interim sanctions. (a) If the Deputy Under Secretary concludes, after issuing a charging letter under § 356.16 and before a final decision is rendered, that interim sanctions are necessary...

  18. An Interim President Sets the Stage

    ERIC Educational Resources Information Center

    Guardo, Carol J.

    2006-01-01

    An interim president often plays a crucial role in leading a college or university. In some instances, the interim can address and resolve troublesome issues and thus clear the way for the new president to generate progress. In others, the interim stays the course so that the institution maintains its momentum and seizes strategic opportunities to…

  19. 29 CFR 1614.505 - Interim relief.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 4 2010-07-01 2010-07-01 false Interim relief. 1614.505 Section 1614.505 Labor Regulations... OPPORTUNITY Remedies and Enforcement § 1614.505 Interim relief. (a)(1) When the agency appeals and the case... interim relief. (2) Service under the temporary or conditional restoration provisions of paragraph...

  20. 22 CFR 127.8 - Interim suspension.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Interim suspension. 127.8 Section 127.8 Foreign... Interim suspension. (a) The Managing Director of the Directorate of Defense Trade Controls or the Director of the Office of Defense Trade Controls Compliance is authorized to order the interim suspension...

  1. 32 CFR 776.82 - Interim suspension.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 5 2010-07-01 2010-07-01 false Interim suspension. 776.82 Section 776.82... Complaint Processing Procedures § 776.82 Interim suspension. (a) Where the Rules Counsel determines there is... interim suspension, pending completion of a professional responsibility investigation. The...

  2. Indiana Corn Dry Mill

    SciTech Connect

    2006-09-01

    The goal of this project is to perform engineering, project design, and permitting for the creation and commercial demonstration of a corn dry mill biorefinery that will produce fuel-grade ethanol, distillers dry grain for animal feed, and carbon dioxide for industrial use.

  3. Alternate Fuels Combustion Research

    DTIC Science & Technology

    1983-10-01

    AFWAL-TR-83-2057 AD A13 8 5 7 5 ALTERNATE FUELS COMBUSTION RESEARCH PHASE RI ’~*~~4 & IWITEY CMAAA * ’s~t:Uwz, ONTARIO October 1983 I•oerls Report...83-2057 P_______________ C TITLE (mod ,,--tt-) 5. TYPE OF REPORT A PERIOD COVERED Alternate Fuels ioahusticn Research Interim Report for Period Phase...I$. KEY WORDS (Continue on reverse sirte it necessear and identify by block number) FUELS ALTERNATE FUELS GAS TURBINE COMBUSTION EXHAUST EMISSIONS 0

  4. Shippingport Spent Fuel Canister System Description

    SciTech Connect

    JOHNSON, D.M.

    2001-06-26

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

  5. Shippingport Spent Fuel Canister System Description

    SciTech Connect

    JOHNSON, D.M.

    2000-03-27

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

  6. Collection Analysis Project. Interim Report.

    ERIC Educational Resources Information Center

    Collier, Carol; And Others

    This interim report of the University of Wyoming's Collection Analysis Project, whose purpose is to examine the university library's collection development process and make workable recommendations for process improvement, consists of a brief history of the library's collection, including a review of the evolution of the collection development…

  7. Data breaches. Interim final rule.

    PubMed

    2007-06-22

    This document establishes regulations to address data breaches regarding sensitive personal information that is processed or maintained by the Department of Veterans Affairs (VA). The regulations implement certain provisions of Title IX of the Veterans Benefits, Health Care, and Information Technology Act of 2006, which require promulgation of these regulations as an interim final rule.

  8. Primer for the Interim Chair

    ERIC Educational Resources Information Center

    Soltys, Stephen M.

    2011-01-01

    Objective: Being successful in the role of an Interim Chair requires an approach to transitional leadership that is different from that of individuals filling the Chair role permanently. This article reviews pertinent literature on the topic. Method: The author reviewed the literature, cited pertinent articles, and supplemented with personal…

  9. Project Canada West. Interim Report.

    ERIC Educational Resources Information Center

    Western Curriculum Project on Canada Studies, Edmonton (Alberta).

    This first interim report of project Canada West, initiated in April, 1970, includes three papers. "Curriculum Development in Urbanization," by Dr. R. H. Sabey, Acting Executive Director, is an overview of the Project's history, purpose, and the 5 year plan for phases of research development. (See SO 000 283 for complete abstract.) Two…

  10. Bases for extrapolating materials durability in fuel storage pools

    SciTech Connect

    Johnson, A.B. Jr.

    1994-12-01

    A major body of evidence indicates that zirconium alloys have the most consistent and reliable durability in wet storage, justifying projections of safe wet storage greater than 50 y. Aluminum alloys have the widest range of durabilities in wet storage; systematic control and monitoring of water chemistry have resulted in low corrosion rates for more than two decades on some fuels and components. However, cladding failures have occurred in a few months when important parameters were not controlled. Stainless steel is extremely durable when stress, metallurgical and water chemistry factors are controlled. LWR SS cladding has survived for 25 y in wet storage. However, sensitized, stressed SS fuels and components have seriously degraded in fuel storage pools (FSPs) at {approximately} 30 C. Satisfactory durability of fuel assembly and FSP component materials in extended wet storage requires investments in water quality management and surveillance, including chemical and biological factors. The key aspect of the study is to provide storage facility operators and other decision makers a basis to judge the durability of a given fuel type in wet storage as a prelude to basing other fuel management plans (e.g. dry storage) if wet storage will not be satisfactory through the expected period of interim storage.

  11. Chemical Engineering Division fuel cycle programs. Quarterly progress report, April-June 1979. [Pyrochemical/dry processing; waste encapsulation in metal; transport in geologic media

    SciTech Connect

    Steindler, M.J.; Ader, M.; Barletta, R.E.

    1980-09-01

    For pyrochemical and dry processing materials development included exposure to molten metal and salt of Mo-0.5% Ti-0.07% Ti-0.01% C, Mo-30% W, SiC, Si/sub 2/ON/sub 2/, ZrB/sub 2/-SiC, MgAl/sub 2/O/sub 4/, Al/sub 2/O/sub 3/, AlN, HfB/sub 2/, Y/sub 2/O/sub 3/, BeO, Si/sub 3/N/sub 4/, nickel nitrate-infiltrated W, W-coated Mo, and W-metallized alumina-yttria. Work on Th-U salt transport processing included solubility of Th in liquid Cd, defining the Cd-Th and Cd-Mg-Th phase diagrams, ThO/sub 2/ reduction experiments, and electrolysis of CaO in molten salt. Work on pyrochemical processes and associated hardware for coprocessing U and Pu in spent FBR fuels included a second-generation computer model of the transport process, turntable transport process design, work on the U-Cu-Mg system, and U and Pu distribution coefficients between molten salt and metal. Refractory metal vessels are being service-life tested. The chloride volatility processing of Th-based fuel was evaluated for its proliferation resistance, and a preliminary ternary phase diagram for the Zn-U-Pu system was computed. Material characterization and process analysis were conducted on the Exportable Pyrochemical process (Pyro-Civex process). Literature data on oxidation of fissile metals to oxides were reviewed. Work was done on chemical bases for the reprocessing of actinide oxides in molten salts. Flowsheets are being developed for the processing of fuel in molten tin. Work on encapsulation of solidified radioactive waste in metal matrix included studies of leach rate of crystalline waste materials and of the impact resistance of metal-matrix waste forms. In work on the transport properties of nuclear waste in geologic media, adsorption of Sr on oolitic limestone was studied, as well as the migration of Cs in basalt. Fitting of data on the adsorption of iodate by hematite to a mathematical model was attempted.

  12. Fuel Lubricity Requirements for Diesel Injection Systems

    DTIC Science & Technology

    1991-02-01

    8. Montemayor , A.F. and Owens, E.C., "Comparison of 6.2L Arctic and Standard Fuel Injection Pumps Using JP-8 Fuel," Interim Report BFLRF No. 218 (AD A...injection pumps. The unitest pump stand and test equipment specification are described in more detail in the report by Montemayor and Owens*. A schematic...operation. * Montemayor , A.F. and Owens, E.C., "Comparison of 6.2L Arctic and Standard Fuel Injection Pumps Using JP-8 Fuel," Interim Report BFLRF No. 218

  13. 41 CFR 101-29.204 - Interim Federal specification.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 41 Public Contracts and Property Management 2 2010-07-01 2010-07-01 true Interim Federal... PRODUCT DESCRIPTIONS 29.2-Definitions § 101-29.204 Interim Federal specification. An interim Federal... agencies. Interim amendments to Federal Specifications and amendments to interim Federal specifications...

  14. MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY

    SciTech Connect

    Hill, Thomas J

    2005-09-01

    The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to

  15. Evaluation of 25-Percent ATJ Fuel Blends in the John Deere 4045HF 280 Engine

    DTIC Science & Technology

    2014-08-01

    UNCLASSIFIED UNCLASSIFIED EVALUATION OF 25-PERCENT ATJ FUEL BLENDS IN THE JOHN DEERE 4045HF280 ENGINE INTERIM REPORT TFLRF No. 458...FUEL BLENDS IN THE JOHN DEERE 4045HF280 ENGINE INTERIM REPORT TFLRF No. 458 by Adam C. Brandt Edwin A. Frame U.S. Army TARDEC Fuels and...Fuel Blends in the John Deere 4045HF280 Engine 5a. CONTRACT NUMBER W56HZV-09-C-0100 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6

  16. 1987 Federal interim storage fee study: A technical and economic analysis

    SciTech Connect

    Not Available

    1987-09-01

    This document is the latest in a series of reports that are published annually by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE). This information in the report, which was prepared by E.R. Johnson Associates under subcontract to PNL, will be used by the DOE to establish a payment schedule for interim storage of spent nuclear fuel under the Federal Interim Storage (FIS) Program, which was mandated by the Nuclear Waste Policy Act of 1982. The information in this report will be used to establish the schedule of charges for FIS services for the year commencing January 1, 1988. 13 tabs.

  17. Burn site groundwater interim measures work plan.

    SciTech Connect

    Witt, Jonathan L.; Hall, Kevin A.

    2005-05-01

    This Work Plan identifies and outlines interim measures to address nitrate contamination in groundwater at the Burn Site, Sandia National Laboratories/New Mexico. The New Mexico Environment Department has required implementation of interim measures for nitrate-contaminated groundwater at the Burn Site. The purpose of interim measures is to prevent human or environmental exposure to nitrate-contaminated groundwater originating from the Burn Site. This Work Plan details a summary of current information about the Burn Site, interim measures activities for stabilization, and project management responsibilities to accomplish this purpose.

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

    SciTech Connect

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

    1982-12-31

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

  19. Managing Spent Nuclear Fuel at the Idaho National Laboratory

    SciTech Connect

    Thomas Hill; Denzel L. Fillmore

    2005-10-01

    The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

  20. Drying Results of K-Basin Damaged/Corroded SNF Internal Sludge and Surface Coating

    SciTech Connect

    Abrefah, J.; Alexander, D.L.; Marschman, S.C.

    2000-09-21

    Experiments were performed using a thermogravimetric analysis (TGA) system by Pacific Northwest National Laboratory (PNNL)to study the drying behavior of the K-Basin spent nuclear fuel (SNF) internal sludge and two different surface coatings of SNF elements. These measurements were conducted in support of the safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor fuel stored at K-Basin to an interim storage facility. These limited experiments on the corrosion products of K-Basin SNF material were part of the broad studies performed to ascertain the bounding pressurization of the Multi-Canister Overpack (MCO). Seven SNF internal sludge samples taken from different damage regions of three damaged/corroded outer K-Basin SNF elements were dried. Additionally, two surface coating samples taken from two SNF elements stored at K-West were tested. All the tests were performed in a vacuum atmosphere with the same temperature ramp rate of about 0.4 C/ min. Each TGA test sample was weighed before and after the test on a balance located in the Shielded Analytical Laboratory hot cell. The test samples were vacuum dried in the TGA system for about 24 hours prior to heating them at the rate of 0.4 C/min. The observations from the weight change data are summarized below.

  1. Dedicated-site, interim storage of high-level nuclear waste as part of the management system

    PubMed Central

    Zen, E-an

    1980-01-01

    Dedicated-site interim storage of high-level reprocessed nuclear waste and of spent fuel rods is proposed as a long-term integral part of the systems approach of the national nuclear waste isolation program. Separation of interim sites for retrievable storage from permanent-disposal repositories should enhance ensurance of the performance of the latter; maintenance of retrievability at separate sites also has many advantages in both safety and possible use of waste as resources. Interim storage sites probably will not be needed beyond about 100 years from now, so the institutional and technical considerations involved in their choice should be much less stringent than those for the selection of permanent sites. Development of interim sites must be concurrent with unabated effort to identify and to develop permanent repositories. PMID:16592904

  2. 45 CFR 1623.6 - Interim funding.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 45 Public Welfare 4 2013-10-01 2013-10-01 false Interim funding. 1623.6 Section 1623.6 Public Welfare Regulations Relating to Public Welfare (Continued) LEGAL SERVICES CORPORATION SUSPENSION PROCEDURES § 1623.6 Interim funding. (a) Pending the completion of suspension proceedings under this...

  3. 45 CFR 1623.6 - Interim funding.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 45 Public Welfare 4 2014-10-01 2014-10-01 false Interim funding. 1623.6 Section 1623.6 Public Welfare Regulations Relating to Public Welfare (Continued) LEGAL SERVICES CORPORATION SUSPENSION PROCEDURES § 1623.6 Interim funding. (a) Pending the completion of suspension proceedings under this...

  4. 45 CFR 1623.6 - Interim funding.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 45 Public Welfare 4 2011-10-01 2011-10-01 false Interim funding. 1623.6 Section 1623.6 Public Welfare Regulations Relating to Public Welfare (Continued) LEGAL SERVICES CORPORATION SUSPENSION PROCEDURES § 1623.6 Interim funding. (a) Pending the completion of suspension proceedings under this...

  5. 45 CFR 1623.6 - Interim funding.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 45 Public Welfare 4 2012-10-01 2012-10-01 false Interim funding. 1623.6 Section 1623.6 Public Welfare Regulations Relating to Public Welfare (Continued) LEGAL SERVICES CORPORATION SUSPENSION PROCEDURES § 1623.6 Interim funding. (a) Pending the completion of suspension proceedings under this...

  6. 40 CFR 94.12 - Interim provisions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Interim provisions. 94.12 Section 94.12 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED... for Compression-Ignition Marine Engines § 94.12 Interim provisions. This section contains...

  7. 45 CFR 1623.6 - Interim funding.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 45 Public Welfare 4 2010-10-01 2010-10-01 false Interim funding. 1623.6 Section 1623.6 Public Welfare Regulations Relating to Public Welfare (Continued) LEGAL SERVICES CORPORATION SUSPENSION PROCEDURES § 1623.6 Interim funding. (a) Pending the completion of suspension proceedings under this...

  8. 40 CFR 1042.145 - Interim provisions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Interim provisions. 1042.145 Section... Standards and Related Requirements § 1042.145 Interim provisions. (a) General. The provisions in this section apply instead of other provisions in this part. This section describes when these...

  9. 39 CFR 952.6 - Interim impounding.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 39 Postal Service 1 2010-07-01 2010-07-01 false Interim impounding. 952.6 Section 952.6 Postal Service UNITED STATES POSTAL SERVICE PROCEDURES RULES OF PRACTICE IN PROCEEDINGS RELATIVE TO FALSE REPRESENTATION AND LOTTERY ORDERS § 952.6 Interim impounding. In preparation for or during the pendency of...

  10. Locating Interim Assessments within Teachers' Assessment Practice

    ERIC Educational Resources Information Center

    Riggan, Matthew; Olah, Leslie Nabors

    2011-01-01

    Promising research on the teaching and learning impact of classroom-embedded formative assessment has spawned interest in a broader array of assessment tools and practices, including interim assessment. Although researchers have begun to explore the impact of interim assessments in the classroom, like other assessment tools and practices, they…

  11. 20 CFR 410.234 - Interim provisions.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 20 Employees' Benefits 2 2010-04-01 2010-04-01 false Interim provisions. 410.234 Section 410.234 Employees' Benefits SOCIAL SECURITY ADMINISTRATION FEDERAL COAL MINE HEALTH AND SAFETY ACT OF 1969, TITLE IV... Evidence § 410.234 Interim provisions. (a) Notwithstanding any other provision of this subpart, a...

  12. 24 CFR 35.820 - Interim controls.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... completion of the risk assessment. In units in which a child of less than 6 years of age moves in after the completion of the risk assessment, interim controls shall be completed no later than 90 days after the move... property, interim controls shall be completed no later than 12 months after completion of the...

  13. 24 CFR 35.820 - Interim controls.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... completion of the risk assessment. In units in which a child of less than 6 years of age moves in after the completion of the risk assessment, interim controls shall be completed no later than 90 days after the move... property, interim controls shall be completed no later than 12 months after completion of the...

  14. 24 CFR 35.820 - Interim controls.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... completion of the risk assessment. In units in which a child of less than 6 years of age moves in after the completion of the risk assessment, interim controls shall be completed no later than 90 days after the move... property, interim controls shall be completed no later than 12 months after completion of the...

  15. 24 CFR 35.820 - Interim controls.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... completion of the risk assessment. In units in which a child of less than 6 years of age moves in after the completion of the risk assessment, interim controls shall be completed no later than 90 days after the move... property, interim controls shall be completed no later than 12 months after completion of the...

  16. 24 CFR 35.1330 - Interim controls.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... Development LEAD-BASED PAINT POISONING PREVENTION IN CERTAIN RESIDENTIAL STRUCTURES Methods and Standards for Lead-Paint Hazard Evaluation and Hazard Reduction Activities § 35.1330 Interim controls. Interim controls of lead-based paint hazards identified in a risk assessment shall be conducted in accordance...

  17. 24 CFR 35.1330 - Interim controls.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... Development LEAD-BASED PAINT POISONING PREVENTION IN CERTAIN RESIDENTIAL STRUCTURES Methods and Standards for Lead-Paint Hazard Evaluation and Hazard Reduction Activities § 35.1330 Interim controls. Interim controls of lead-based paint hazards identified in a risk assessment shall be conducted in accordance...

  18. 24 CFR 35.1330 - Interim controls.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... Development LEAD-BASED PAINT POISONING PREVENTION IN CERTAIN RESIDENTIAL STRUCTURES Methods and Standards for Lead-Paint Hazard Evaluation and Hazard Reduction Activities § 35.1330 Interim controls. Interim controls of lead-based paint hazards identified in a risk assessment shall be conducted in accordance...

  19. 24 CFR 35.1330 - Interim controls.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... Development LEAD-BASED PAINT POISONING PREVENTION IN CERTAIN RESIDENTIAL STRUCTURES Methods and Standards for Lead-Paint Hazard Evaluation and Hazard Reduction Activities § 35.1330 Interim controls. Interim controls of lead-based paint hazards identified in a risk assessment shall be conducted in accordance...

  20. 24 CFR 35.1330 - Interim controls.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... Development LEAD-BASED PAINT POISONING PREVENTION IN CERTAIN RESIDENTIAL STRUCTURES Methods and Standards for Lead-Paint Hazard Evaluation and Hazard Reduction Activities § 35.1330 Interim controls. Interim controls of lead-based paint hazards identified in a risk assessment shall be conducted in accordance...

  1. Solid waste burial grounds interim safety analysis

    SciTech Connect

    Saito, G.H.

    1994-10-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

  2. Fusion Breeder Program interim report

    SciTech Connect

    Moir, R.; Lee, J.D.; Neef, W.

    1982-06-11

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83.

  3. Vet Centers. Interim final rule.

    PubMed

    2015-08-04

    The Department of Veterans Affairs (VA) is amending its medical regulation that governs Vet Center services. The National Defense Authorization Act for Fiscal Year 2013 (the 2013 Act) requires Vet Centers to provide readjustment counseling services to broader groups of veterans, members of the Armed Forces, including a member of a reserve component of the Armed Forces, and family members of such veterans and members. This interim final rule amends regulatory criteria to conform to the 2013 Act, to include new and revised definitions.

  4. In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

    SciTech Connect

    Gauld, Ian C.; Hu, Jianwei; De Baere, P.; Vaccaro, S.; Schwalbach, P.; Liljenfeldt, Henrik; Tobin, Stephen

    2015-01-01

    Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel

  5. Cold vacuum drying facility design requirements

    SciTech Connect

    IRWIN, J.J.

    1999-07-01

    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.

  6. Numerical Estimation of the Spent Fuel Ratio

    SciTech Connect

    Lindgren, Eric R.; Durbin, Samuel; Wilke, Jason; Margraf, J.; Dunn, T. A.

    2016-01-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  7. 40 CFR 270.73 - Termination of interim status.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Termination of interim status. 270.73... (CONTINUED) EPA ADMINISTERED PERMIT PROGRAMS: THE HAZARDOUS WASTE PERMIT PROGRAM Interim Status § 270.73 Termination of interim status. Interim status terminates when: (a) Final administrative disposition of...

  8. 46 CFR 309.101 - Amendment of interim binders.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 8 2010-10-01 2010-10-01 false Amendment of interim binders. 309.101 Section 309.101... INSURANCE § 309.101 Amendment of interim binders. The interim binder for a vessel whose stated valuation is... after the attachment of the insurance under the interim binder to which such valuation...

  9. 40 CFR 270.71 - Operation during interim status.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Operation during interim status. 270... (CONTINUED) EPA ADMINISTERED PERMIT PROGRAMS: THE HAZARDOUS WASTE PERMIT PROGRAM Interim Status § 270.71 Operation during interim status. (a) During the interim status period the facility shall not: (1)...

  10. 49 CFR 106.35 - Interim final rule.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Interim final rule. 106.35 Section 106.35... PHMSA Rulemaking Documents § 106.35 Interim final rule. An interim final rule is issued without first... requirements and their effective date. PHMSA may issue an interim final rule if it finds, for good cause,...

  11. 13 CFR 120.890 - Source of interim financing.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Source of interim financing. 120... Development Company Loan Program (504) Interim Financing § 120.890 Source of interim financing. A Project may use interim financing for all Project costs except the Borrower's contribution. Any source...

  12. 5 CFR 772.102 - Interim personnel actions.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 5 Administrative Personnel 2 2010-01-01 2010-01-01 false Interim personnel actions. 772.102... REGULATIONS (CONTINUED) INTERIM RELIEF General § 772.102 Interim personnel actions. When an employee or... administrative judge granting interim relief under 5 U.S.C. 7701(b)(2)(A) and a petition for review of...

  13. 33 CFR 1.05-45 - Interim rule.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Interim rule. 1.05-45 Section 1... PROVISIONS Rulemaking § 1.05-45 Interim rule. (a) An interim rule may be issued when it is in the public... example, an interim rule may be issued in instances when normal procedures for notice and comment prior...

  14. Spent fuel management status perspectives in Korea

    SciTech Connect

    Park, H.S.; Lee, J.S.; Kim, B.T. )

    1992-01-01

    Concomitant with steadily increasing nuclear power program in Korea, a national radioactive waste management program has been in initial implementation stage for several years. In late 1990, however, a serious confrontation was witnessed at Anmyon area where residents expressed strong opposition against any possibility to consider that site as a potential candidate for waste disposal by the Authority. As far as spent fuel management is concerned, an interim storage policy was adopted by Korean Atomic Energy Commission. A decision to build a centralized wet storage facility was made followed by a conceptual design. Due to the incident at Anmyon site, the public has became more concerned about radioactive wastes management. Parallel efforts are being made to ameliorate public acceptance in regard to radioactive waste management and in particular to spent fuel management. There are substantial uncertainties, however, whether any site could be found given that precarious mood has been prevailing against radioactive wastes throughout the world. In the meantime waiting for successful siting, various research and development for future perspectives are in order. Of particular importance in such endeavor is to provide technological impetus for future perspectives as well as public acceptance through safety demonstrations of certain viable technology alternatives. The dry storage option, for instance, is acclaimed for intrinsic safety and lower cost as prospective alternative. Combined with rod consolidation, dry storage technologies which have not extensively applied in the past, could be considered as a technological basis for longer term management of spent fuel. Conscious of such global trend, some appropriate programs in preparation for such perspectives have been launched by KAERI.

  15. Effectiveness of Additives in Improving Fuel Lubricity and Preventing Pump Failure at High Temperature

    DTIC Science & Technology

    2013-01-01

    the specified 1,700 rpm, the housing pressure on pump No. 1 dropped significantly. Fuel began spewing out of test stand gear box and the pump seized...UNCLASSIFIED EFFECTIVENESS OF ADDITIVES IN IMPROVING FUEL LUBRICITY AND PREVENTING PUMP FAILURE AT HIGH TEMPERATURE INTERIM REPORT TFLRF...UNCLASSIFIED EFFECTIVENESS OF ADDITIVES IN IMPROVING FUEL LUBRICITY AND PREVENTING PUMP FAILURE AT HIGH TEMPERATURE INTERIM REPORT TFLRF No. 437

  16. Receipt and Storage Issues at the TMI-2 Irradiated Fuel Storage Installation

    SciTech Connect

    Christensen, Allan B.; Custer, Kenneth; Gardner, Rick; Kaylor, James; Stalnaker, James

    2002-07-01

    In less than a year, up to 12 canisters of TMI-2 reactor fuel debris were loaded into each of 28 Dry Storage Containers (DSCs), and placed into interim storage at an Irradiated Spent Fuel Storage Facility (ISFSI) at the Idaho National Engineering and Environmental Laboratory (INEEL). Draining and drying the canisters, loading and welding the DSCs, shipping the DSCs 25 miles, and storing in the ISFSI initially required up to 3 weeks per DSC. Significant time efficiencies were achieved during the early stages, reducing the time to less than one week per DSC. These efficiencies were achieved mostly in canister draining and drying and DSC lid welding, and despite several occurrences that had to be resolved before continuing work. The ISFSI has been operated without issue since, with the exception that license basis monitoring has indicated an unusual pattern of season- and position-dependent hydrogen generation. This paper discusses some of the innovations and storage experiences for the first ISFSI designed for the storage of severely defected fuel. (authors)

  17. EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

    SciTech Connect

    Matloch, L.; Vaccaro, S.; Couland, M.; De Baere, P.; Schwalbach, P.

    2015-07-01

    The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction of encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)

  18. Fuels from Recycling Systems

    ERIC Educational Resources Information Center

    Tillman, David A.

    1975-01-01

    Three systems, operating at sufficient scale, produce fuels that may be alternatives to oil and gas. These three recycling systems are: Black Clawson Fiberclaim, Franklin, Ohio; Union Carbide, South Charleston, West Virginia; and Union Electric, St. Louis, Missouri. These produce a wet fuel, a pyrolytic gas, and a dry fuel, respectively. (BT)

  19. High Temperature Materials Interim Data Qualification Report

    SciTech Connect

    Nancy Lybeck

    2010-08-01

    ABSTRACT Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim FY2010 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under NQA-1 guidelines, and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from two test series within the High Temperature Materials data stream have been entered into the NDMAS vault: 1. Tensile Tests for Sm (i.e., Allowable Stress) Confirmatory Testing – 1,403,994 records have been inserted into the NDMAS database. Capture testing is in process. 2. Creep-Fatigue Testing to Support Determination of Creep-Fatigue Interaction Diagram – 918,854 records have been processed and inserted into the NDMAS database. Capture testing is in process.

  20. Interim Policy for Evaluation of Stereoisomeric Pesticides

    EPA Pesticide Factsheets

    An interim approach for determining data requirements for non-racemic mixtures of stereoisomeric pesticides. These data are needed in order to assess the risk posed to ecosystems and drinking water sources by these mixtures.

  1. 40 CFR 180.319 - Interim tolerances.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... TOLERANCES AND EXEMPTIONS FOR PESTICIDE CHEMICAL RESIDUES IN FOOD Specific Tolerances § 180.319 Interim... pesticide chemicals in or on the following raw agricultural commodities: Substances Uses Tolerance in...

  2. Interim Prosthesis Options for Dental Implants.

    PubMed

    Siadat, Hakimeh; Alikhasi, Marzieh; Beyabanaki, Elaheh

    2016-01-24

    Dental implants have become a popular treatment modality for replacing missing teeth. In this regard, the importance of restoring patients with function during the implant healing period has grown in recent decades. Esthetic concerns, especially in the anterior region of the maxilla, should also be considered until the definitive restoration is delivered. Another indication for such restorations is maintenance of the space required for esthetic and functional definitive restorations in cases where the implant site is surrounded by natural teeth. Numerous articles have described different types of interim prostheses and their fabrication techniques. This article aims to briefly discuss all types of implant-related interim prostheses by different classification including provisional timing (before implant placement, after implant placement in unloading and loading periods), materials, and techniques used for making the restorations, the type of interim prosthesis retention, and definitive restoration. Furthermore, the abutment torque for such restorations and methods for transferring the soft tissue from interim to definitive prostheses are addressed.

  3. Interim-Night Integrated Goggle Head Tracking System (I-Nights). Volume 1. Ground Test Summary

    DTIC Science & Technology

    1992-08-01

    3 Figure 2. Honeywell I-NIGHTS System Drawing . . . . . . . . 4 Figure 3. Kaiser I-NIGHTS System Drawing . . . . . . . . . . 5 Figure 4. Spinal ...thoracolumbar spinal fracture during ejection seat use. The USAF use of the DRI to evaluate ejection seats is embodied in Military Specification: Seat Systems...probability of spinal injury. But, what WT/CG/force combinations are reasonably acceptable during ejection? The I-NIGHTS program helped establish an interim

  4. Seismic Performance of Dry Casks Storage for Long- Term Exposure

    SciTech Connect

    Ibarra, Luis; Sanders, David; Yang, Haori; Pantelides, Chris

    2016-12-30

    The main goal of this study is to evaluate the long-term seismic performance of freestanding and anchored Dry Storage Casks (DSCs) using experimental tests on a shaking table, as well as comprehensive numerical evaluations that include the cask-pad-soil system. The study focuses on the dynamic performance of vertical DSCs, which can be designed as free-standing structures resting on a reinforced concrete foundation pad, or casks anchored to a foundation pad. The spent nuclear fuel (SNF) at nuclear power plants (NPPs) is initially stored in fuel-storage pools to control the fuel temperature. After several years, the fuel assemblies are transferred to DSCs at sites contiguous to the plant, known as Interim Spent Fuel Storage Installations (ISFSIs). The regulations for these storage systems (10 CFR 72) ensure adequate passive heat removal and radiation shielding during normal operations, off-normal events, and accident scenarios. The integrity of the DSCs is important, even if the overpack does not breach, because eventually the spent fuel-rods need to be shipped either to a reprocessing plant or a repository. DSCs have been considered as a temporary storage solution, and usually are licensed for 20 years, although they can be relicensed for operating periods of up to 60 years. In recent years, DSCs have been reevaluated as a potential mid-term solution, in which the operating period may be extended for up to 300 years. At the same time, recent seismic events have underlined the significant risks DSCs are exposed. The consideration of DCSs for storing spent fuel for hundreds of years has created new challenges. In the case of seismic hazard, longer-term operating periods not only lead to larger horizontal accelerations, but also increase the relative effect of vertical accelerations that usually are disregarded for smaller seismic events. These larger seismic demands could lead to casks sliding and tipping over, impacting the concrete pad or adjacent casks. The casks

  5. 78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-13

    ... COMMISSION Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors Reviews AGENCY: Nuclear Regulatory Commission. ACTION: Draft Interim Staff Guidance; request for comment... comment on, draft Interim Staff Guidance (ISG) ESP/COL-ISG-027, ``Interim Staff Guidance...

  6. Rejuvenation of automotive fuel cells

    SciTech Connect

    Kim, Yu Seung; Langlois, David A.

    2016-08-23

    A process for rejuvenating fuel cells has been demonstrated to improve the performance of polymer exchange membrane fuel cells with platinum/ionomer electrodes. The process involves dehydrating a fuel cell and exposing at least the cathode of the fuel cell to dry gas (nitrogen, for example) at a temperature higher than the operating temperature of the fuel cell. The process may be used to prolong the operating lifetime of an automotive fuel cell.

  7. Dry Mouth

    MedlinePlus

    ... or chewing tobacco can increase dry mouth symptoms. Methamphetamine use. Methamphetamine use can cause severe dry mouth and damage to teeth, a condition also known as "meth mouth." If you don't have enough saliva ...

  8. Dry Mouth

    MedlinePlus

    Dry mouth is the feeling that there is not enough saliva in your mouth. Everyone has a dry mouth once in a while - if they are nervous, ... under stress. But if you have a dry mouth all or most of the time, it can ...

  9. Lignite Fuel Enhancement

    SciTech Connect

    Charles Bullinger

    2006-04-03

    This 7th quarterly Technical Progress Report for the Lignite Fuel Enhancement Project summarizes activities from January 1st through March 31st of 2006. It also summarizes the subsequent purchasing activity, dryer/process construction, and testing. The Design Team began conferencing again as construction completed and the testing program began. Primary focus this quarter was construction/installation completion. Phase 1 extension recommendation, and subsequent new project estimate, Forms 424 and 4600 were accepted by DOE headquarters. DOE will complete the application and amended contract. All major mechanical equipment was run, checked out, and tested this quarter. All water, air, and coal flow loops were run and tested. The system was run on January 30th, shut down to adjust equipment timing in the control system on the 31st, and run to 75 ton//hour on February 1st. It ran for seven to eight hours per day until March 20th when ''pairs'' testing ( 24 hour running) began. ''Pairs'' involves comparative testing of unit performance with seven ''wet'' pulverizers versus six ''wet'' and one ''dry''. During the interim, more operators were brought up to speed on system operation and control was shifted to the main Unit No.2 Control Room. The system is run now from the Unit control board operator and an equipment operator checks the system during regular rounds or when an alarm needs verification. The flawless start-up is unprecedented in the industry and credit should be made to the diligence and tenacity of Coal Creek maintenance/checkout staff. Great River Energy and Headwaters did not meet to discuss the Commercialization Plan this quarter. The next meeting is pending data from the drying system. Discussions with Basin Electric, Otter Tail, and Dairyland continue and confidentiality secured as we promote dryers in their stations. Lighting and fire protection were completed in January. Invoices No.12 through No.20 are completed and forwarded following preliminary

  10. Compton Dry-Cask Imaging System

    ScienceCinema

    None

    2016-07-12

    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/

  11. Compton Dry-Cask Imaging System

    SciTech Connect

    2011-01-01

    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/

  12. Methods Data Qualification Interim Report

    SciTech Connect

    R. Sam Alessi; Tami Grimmett; Leng Vang; Dave McGrath

    2010-09-01

    The overall goal of the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) is to maintain data provenance for all NGNP data including the Methods component of NGNP data. Multiple means are available to access data stored in NDMAS. A web portal environment allows users to access data, view the results of qualification tests and view graphs and charts of various attributes of the data. NDMAS also has methods for the management of the data output from VHTR simulation models and data generated from experiments designed to verify and validate the simulation codes. These simulation models represent the outcome of mathematical representation of VHTR components and systems. The methods data management approaches described herein will handle data that arise from experiment, simulation, and external sources for the main purpose of facilitating parameter estimation and model verification and validation (V&V). A model integration environment entitled ModelCenter is used to automate the storing of data from simulation model runs to the NDMAS repository. This approach does not adversely change the why computational scientists conduct their work. The method is to be used mainly to store the results of model runs that need to be preserved for auditing purposes or for display to the NDMAS web portal. This interim report demonstrates the currently development of NDMAS for Methods data and discusses data and its qualification that is currently part of NDMAS.

  13. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    SciTech Connect

    Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel.

  14. Examination of the surface coating removed from K-East Basin fuel elements

    SciTech Connect

    Abrefah, J.; Marschman, S.C.; Jenson, E.D.

    1998-05-01

    This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford site. Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements. Altogether, five samples of coating materials were analyzed. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage.

  15. 46 CFR 309.101 - Amendment of interim binders.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 8 2011-10-01 2011-10-01 false Amendment of interim binders. 309.101 Section 309.101... INSURANCE § 309.101 Amendment of interim binders. The interim binder for a vessel whose stated valuation is established pursuant to this part shall be deemed to have been amended on the first day of the...

  16. 46 CFR 309.101 - Amendment of interim binders.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 8 2013-10-01 2013-10-01 false Amendment of interim binders. 309.101 Section 309.101... INSURANCE § 309.101 Amendment of interim binders. The interim binder for a vessel whose stated valuation is established pursuant to this part shall be deemed to have been amended on the first day of the...

  17. 18 CFR 385.1113 - Interim relief (Rule 1113).

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Interim relief (Rule... Under the NGPA § 385.1113 Interim relief (Rule 1113). (a) The petitioner may at any time file a request for interim relief in a proceeding under this subpart, setting forth the legal and factual basis...

  18. 24 CFR 115.203 - Interim certification procedures.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Interim certification procedures... Certification of Substantially Equivalent Agencies § 115.203 Interim certification procedures. (a) Upon receipt of a request for interim certification filed under § 115.202, the Assistant Secretary may...

  19. 48 CFR 801.690-8 - Interim appointment provisions.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Interim appointment... Authority, and Responsibilities 801.690-8 Interim appointment provisions. (a) To ensure availability of procurement support, an interim appointment may be granted for a limited period of time when a candidate...

  20. 10 CFR 205.288 - Interim and ancillary orders.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 3 2010-01-01 2010-01-01 false Interim and ancillary orders. 205.288 Section 205.288... Distribution of Refunds § 205.288 Interim and ancillary orders. The Director of the Office of Hearings and Appeals or his designee may issue any interim or ancillary orders, or make any rulings or...

  1. 78 FR 49782 - Interim Staff Guidance on Changes During Construction

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-15

    ... COMMISSION Interim Staff Guidance on Changes During Construction AGENCY: Nuclear Regulatory Commission. ACTION: Draft interim staff guidance; request for comment. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing this notice for use of, and to solicit public comment on the draft Interim...

  2. 47 CFR 51.611 - Interim wholesale rates.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 47 Telecommunication 3 2010-10-01 2010-10-01 false Interim wholesale rates. 51.611 Section 51.611... Resale § 51.611 Interim wholesale rates. (a) If a state commission cannot, based on the information... commission may elect to establish an interim wholesale rate as described in paragraph (b) of this section....

  3. An Approach for Evaluating the Technical Quality of Interim Assessments

    ERIC Educational Resources Information Center

    Li, Ying; Marion, Scott; Perie, Marianne; Gong, Brian

    2010-01-01

    Increasing numbers of schools and districts have expressed interest in interim assessment systems to prepare for summative assessments and to improve teaching and learning. However, with so many commercial interim assessments available, schools and districts are struggling to determine which interim assessment is most appropriate to their needs.…

  4. 5 CFR 1209.11 - Duration of stay; interim compliance.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 5 Administrative Personnel 3 2010-01-01 2010-01-01 false Duration of stay; interim compliance... WHISTLEBLOWING Stay Requests § 1209.11 Duration of stay; interim compliance. (a) Duration of stay. A stay becomes...) Interim compliance. An agency must immediately comply with an order granting a stay request. Although...

  5. 40 CFR 270.70 - Qualifying for interim status.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Qualifying for interim status. 270.70... (CONTINUED) EPA ADMINISTERED PERMIT PROGRAMS: THE HAZARDOUS WASTE PERMIT PROGRAM Interim Status § 270.70 Qualifying for interim status. (a) Any person who owns or operates an “existing HWM facility” or a...

  6. 40 CFR 66.23 - Interim recalculation of penalty.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Interim recalculation of penalty. 66... Penalties § 66.23 Interim recalculation of penalty. (a) The Administrator, upon concluding that a previously... respond in accordance with the provisions of § 66.51. The decision to accept the interim calculation or...

  7. 12 CFR 541.18 - Interim Federal savings association.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Interim Federal savings association. 541.18... REGULATIONS AFFECTING FEDERAL SAVINGS ASSOCIATIONS § 541.18 Interim Federal savings association. The term interim Federal savings association means a Federal savings association chartered by the Office...

  8. 12 CFR 541.19 - Interim state savings association.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Interim state savings association. 541.19... REGULATIONS AFFECTING FEDERAL SAVINGS ASSOCIATIONS § 541.19 Interim state savings association. The term interim state savings association means a savings association, other than a Federal savings...

  9. 48 CFR 301.603-3 - Interim appointments.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Interim appointments. 301... Interim appointments. If it is essential to appoint an individual as a Contracting Officer who does not... HCA (non-delegable) may make an interim appointment for up to 2 years. If an extension of time...

  10. 42 CFR 417.570 - Interim per capita payments.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 42 Public Health 3 2010-10-01 2010-10-01 false Interim per capita payments. 417.570 Section 417... PREPAYMENT PLANS Medicare Payment: Cost Basis § 417.570 Interim per capita payments. (a) Principle of payment. (1) CMS makes monthly advance payments equivalent to the HMO's or CMP's interim per capita rate...

  11. 5 CFR 531.414 - Interim within-grade increase.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 5 Administrative Personnel 1 2010-01-01 2010-01-01 false Interim within-grade increase. 531.414... UNDER THE GENERAL SCHEDULE Within-Grade Increases § 531.414 Interim within-grade increase. (a) An interim within-grade increase shall be granted to an employee who has: (1) Appealed a negative...

  12. 49 CFR 37.193 - Interim service requirements.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 1 2010-10-01 2010-10-01 false Interim service requirements. 37.193 Section 37... WITH DISABILITIES (ADA) Over-the-Road Buses (OTRBs) § 37.193 Interim service requirements. (a) Until... of accessible OTRBs, the operator shall meet the following interim service requirements:...

  13. 28 CFR 94.41 - Interim emergency payment.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Interim emergency payment. 94.41 Section... Terrorism Victim Expense Reimbursement Program Payment of Claims § 94.41 Interim emergency payment. Claimants may apply for an interim emergency payment, prior to a determination under § 94.21(a). If...

  14. 41 CFR 101-29.206 - Interim Federal standard.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 41 Public Contracts and Property Management 2 2010-07-01 2010-07-01 true Interim Federal standard. 101-29.206 Section 101-29.206 Public Contracts and Property Management Federal Property Management... DESCRIPTIONS 29.2-Definitions § 101-29.206 Interim Federal standard. An interim Federal standard is a...

  15. 24 CFR 115.202 - Request for interim certification.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Request for interim certification... Certification of Substantially Equivalent Agencies § 115.202 Request for interim certification. (a) A request for interim certification under this subpart shall be filed with the Assistant Secretary by the...

  16. 40 CFR 270.72 - Changes during interim status.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Changes during interim status. 270.72... (CONTINUED) EPA ADMINISTERED PERMIT PROGRAMS: THE HAZARDOUS WASTE PERMIT PROGRAM Interim Status § 270.72 Changes during interim status. (a) Except as provided in paragraph (b), the owner or operator of...

  17. 7 CFR 900.70 - Applications for interim relief.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 8 2010-01-01 2010-01-01 false Applications for interim relief. 900.70 Section 900.70... Applications for interim relief. (a) Filing the application. A person who has filed a petition pursuant to... statement of the objections, if any, of the Administrator to the application for interim relief, and may...

  18. 40 CFR 264.3 - Relationship to interim status standards.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Relationship to interim status... FACILITIES General § 264.3 Relationship to interim status standards. A facility owner or operator who has fully complied with the requirements for interim status—as defined in section 3005(e) of RCRA...

  19. 5 CFR 772.102 - Interim personnel actions.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 5 Administrative Personnel 2 2011-01-01 2011-01-01 false Interim personnel actions. 772.102 Section 772.102 Administrative Personnel OFFICE OF PERSONNEL MANAGEMENT (CONTINUED) CIVIL SERVICE REGULATIONS (CONTINUED) INTERIM RELIEF General § 772.102 Interim personnel actions. When an employee...

  20. 5 CFR 772.102 - Interim personnel actions.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 5 Administrative Personnel 2 2013-01-01 2013-01-01 false Interim personnel actions. 772.102 Section 772.102 Administrative Personnel OFFICE OF PERSONNEL MANAGEMENT (CONTINUED) CIVIL SERVICE REGULATIONS (CONTINUED) INTERIM RELIEF General § 772.102 Interim personnel actions. When an employee...

  1. 5 CFR 772.102 - Interim personnel actions.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 5 Administrative Personnel 2 2012-01-01 2012-01-01 false Interim personnel actions. 772.102 Section 772.102 Administrative Personnel OFFICE OF PERSONNEL MANAGEMENT (CONTINUED) CIVIL SERVICE REGULATIONS (CONTINUED) INTERIM RELIEF General § 772.102 Interim personnel actions. When an employee...

  2. CMM Interim Check Design of Experiments (U)

    SciTech Connect

    Montano, Joshua Daniel

    2015-07-29

    Coordinate Measuring Machines (CMM) are widely used in industry, throughout the Nuclear Weapons Complex and at Los Alamos National Laboratory (LANL) to verify part conformance to design definition. Calibration cycles for CMMs at LANL are predominantly one year in length and include a weekly interim check to reduce risk. The CMM interim check makes use of Renishaw’s Machine Checking Gauge which is an off-the-shelf product simulates a large sphere within a CMM’s measurement volume and allows for error estimation. As verification on the interim check process a design of experiments investigation was proposed to test a couple of key factors (location and inspector). The results from the two-factor factorial experiment proved that location influenced results more than the inspector or interaction.

  3. EMCS Retrofit Analysis - Interim Report

    SciTech Connect

    Diamond, R.C.; Salsbury, T.I.; Bell, G.C.; Huang, Y.J.; Sezgen, A.O.; Mazzucchi, R.; Romberger, J.

    1999-03-01

    This report presents the interim results of analyses carried out in the Phillip Burton Federal Building in San Francisco from 1996 to 1998. The building is the site of a major demonstration of the BACnet communication protocol. The energy management and control systems (EMCS) in the building were retrofitted with BACnet compatible controllers in order to integrate certain existing systems on one common network. In this respect, the project has been a success. Interoperability of control equipment from different manufacturers has been demonstrated in a real world environment. Besides demonstrating interoperability, the retrofits carried out in the building were also intended to enhance control strategies and capabilities, and to produce energy savings. This report presents analyses of the energy usage of HVAC systems in the building, control performance, and the reaction of the building operators. The report does not present an evaluation of the performance capabilities of the BACnet protocol. A monitoring system was installed in the building that parallels many of the EMCS sensors and data were archived over a three-year period. The authors defined pre-retrofit and post-retrofit periods and analyzed the corresponding data to establish the changes in building performance resulting from the retrofit activities. The authors also used whole-building energy simulation (DOE-2) as a tool for evaluating the effect of the retrofit changes. The results of the simulation were compared with the monitored data. Changes in operator behavior were assessed qualitatively with questionnaires. The report summarizes the findings of the analyses and makes several recommendations as to how to achieve better performance. They maintain that the full potential of the EMCS and associated systems is not being realized. The reasons for this are discussed along with possible ways of addressing this problem. They also describe a number of new technologies that could benefit systems of the type

  4. Fuel cell systems program plan: Fiscal year 1988

    NASA Astrophysics Data System (ADS)

    1988-05-01

    The DOE fuel cell program supports high-risk, high-payoff technology development. This provides industry with the capability to develop and apply fuel cell systems that use conventional and alternative hydrocarbon fuels. The principal DOE fuel cell program goal is to develop cost effective, efficient, and environmentally benign fuel cell systems which will operate on coal-based fuels (and dual fuels: coal and gas) in multiple end use sectors. In the near-term, and as an interim step in achieving this goal, distillate fuel and natural fuel cell system technologies will also be developed. This interim step is a logical progression to the more complex coal-fueled systems and provides a marketable gas-fueled technology. The specific near- to mid-term (mid-1990s) objectives are to develop the key fuel cell technology for phosphoric acid, molten carbonate, and solid oxide fuel cell systems, and to evaluate and conduct research on more advanced fuel cell concepts that would further improve technical and economic performance. Advanced fuel cell concepts may in the long-term (post 2000) offer potential for use in additional applications such as in the transportation and residential sectors. In these applications oil could be displaced by fuel cell systems using fuels derived from coal.

  5. Arrival condition of spent fuel after storage, handling, and transportation

    SciTech Connect

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables.

  6. Criticality safety evaluation report for FFTF 42% fuel assemblies

    SciTech Connect

    Richard, R.F.

    1997-10-28

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC).

  7. Interim Control Module Extraction Strut Assembly

    NASA Technical Reports Server (NTRS)

    Hansen, Chris

    1999-01-01

    The Interim Control Module is a propulsion and attitude control module for the International Space Station developed to serve as a replacement for the Russian Service Module in the event that Russia does not deliver the hardware as expected. The hardware was developed through a joint NASA/Naval Research Laboratory program and is based on a Department of Defense payload. This paper discusses the development and testing of the Extraction Strut hardware that will be used to assist in removing the Interim Control Module from the Shuttle's payload bay.

  8. Interim Land Use Control Implementation Plan

    NASA Technical Reports Server (NTRS)

    Applegate, Joseph L.

    2014-01-01

    This Interim Land Use Control Implementation Plan (LUCIP) has been prepared to inform current and potential future users of the Kennedy Space Center (KSC) Contractors Road Heavy Equipment (CRHE) Area (SWMU 055; "the Site") of institutional controls that have been implemented at the Site1. Although there are no current unacceptable risks to human health or the environment associated with the CRHE Area, an interim institutional land use control (LUC) is necessary to prevent human health exposure to volatile organic compound (VOC)-affected groundwater at the Site. Controls will include periodic inspection, condition certification, and agency notification.

  9. Debate heats up over potential Interim Nuclear Waste Repository, as studies of Yucca Mountain continue

    NASA Astrophysics Data System (ADS)

    Showstack, Randy

    With spent nuclear fuel piling up at power plants around the United States, and with a potential permanent nuclear waste repository at Nevada's Yucca Mountain not scheduled to accept waste until 11 years from now in the year 2010, the nuclear energy industry and many members of Congress have renewed their push to establish an interim repository at the adjacent Nevada Test Site of nuclear bombs.At a sometimes contentious March 12 hearing to consider the Nuclear Waste Policy Act of 1999 (House Resolution 45) that would require an interim facility to begin accepting waste in 2003, bill cosponsor Rep. Jim Barton (R-Tex.) told Energy Secretary Bill Richardson that he preferred that Congress and the Clinton Administration negotiate rather than fight over the measure.

  10. Fuel flexible fuel injector

    DOEpatents

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  11. Co-production of activated carbon, fuel-gas, and oil from the pyrolysis of corncob mixtures with wet and dried sewage sludge.

    PubMed

    Shao, Linlin; Jiang, Wenbo; Feng, Li; Zhang, Liqiu

    2014-06-01

    This study explored the amount and composition of pyrolysis gas and oil derived from wet material or dried material during the preparation of sludge-corncob activated carbon, and evaluated the physicochemical and surface properties of the obtained two types of sludge-corncob-activated carbons. For wet material, owing to the presence of water, the yields of sludge-corncob activated carbon and the oil fraction slightly decreased while the yield of gases increased. The main pyrolysis gas compounds were H2 and CO2, and more H2 was released from wet material than dried material, whereas the opposite holds for CO2 Heterocyclics, nitriles, organic acids, and steroids were the major components of pyrolysis oil. Furthermore, the presence of water in wet material reduced the yield of polycyclic aromatic hydrocarbons from 6.76% to 5.43%. The yield of furfural, one of heterocyclics, increased sharply from 3.51% to 21.4%, which could be explained by the enhanced hydrolysis of corncob. In addition, the surface or chemical properties of the two sludge-corncob activated carbons were almost not affected by the moisture content of the raw material, although their mesopore volume and diameter were different. In addition, the adsorption capacities of the two sludge-corncob activated carbons towards Pb and nitrobenzene were nearly identical.

  12. Comparing hydrogen and hydrocarbon booster fuels

    NASA Technical Reports Server (NTRS)

    Martin, James A.

    1988-01-01

    The present evaluation of the consequences of hydrogen and hydrocarbon fuels as the basis of launch vehicle booster rocket-stage performance notes that hydrocarbon fuels lead to lower vehicle dry mass, for low-velocity requirements, while hydrogen fuel furnishes lower dry mass. Vehicles employing both types of fuel attempt to take advantage of the low intercept and slope of hydrocarbon fuel at low velocity, and subsequently, of the slope of the hydrogen curves at higher velocities.

  13. Technology for Use of Variable Quality Fuels

    DTIC Science & Technology

    1987-12-01

    AD-A 198 782 C TECHNOLOGY FOR USE OF VARIABLE QUALITY FUELS INTERIM REPORT BFLRF No. 239 By J.N. Bowden D.M. Yost 6 A.F. Montemayor L.L. Stavinoha...Operator’s Manual (Crew) for Armored Reconnaissance/Airborne Assault Vehicle, 10 June 1976. 6. Montemayor , A.F., Owens, E.C., "Fuel Property Effects on

  14. LANDFILL BIOREACTOR PERFORMANCE, SECOND INTERIM REPORT

    EPA Science Inventory

    A bioreactor landfill is a landfill that is operated in a manner that is expected to increase the rate and extent of waste decomposition, gas generation, and settlement compared to a traditional landfill. This Second Interim Report was prepared to provide an interpretation of fie...

  15. The Interim Superintendent: A Case Study

    ERIC Educational Resources Information Center

    Bigham, Gary; Nix, Susan J.

    2011-01-01

    Considering the vitally important role that the superintendent plays in the overall functioning and wellbeing of any school district, the filling of that position should never be done in haste. Due to the importance of this process and the time it requires, school districts often employ an interim superintendent. In this single case study, one…

  16. 45 CFR 83.20 - Interim procedures.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 45 Public Welfare 1 2010-10-01 2010-10-01 false Interim procedures. 83.20 Section 83.20 Public Welfare DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL ADMINISTRATION REGULATION FOR THE ADMINISTRATION AND ENFORCEMENT OF SECTIONS 799A AND 845 OF THE PUBLIC HEALTH SERVICE ACT Procedures § 83.20...

  17. 19 CFR 356.18 - Interim sanctions.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 19 Customs Duties 3 2013-04-01 2013-04-01 false Interim sanctions. 356.18 Section 356.18 Customs Duties INTERNATIONAL TRADE ADMINISTRATION, DEPARTMENT OF COMMERCE PROCEDURES AND RULES FOR IMPLEMENTING ARTICLE 1904 OF THE NORTH AMERICAN FREE TRADE AGREEMENT Violation of a Protective Order or a...

  18. 19 CFR 356.18 - Interim sanctions.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 19 Customs Duties 3 2011-04-01 2011-04-01 false Interim sanctions. 356.18 Section 356.18 Customs Duties INTERNATIONAL TRADE ADMINISTRATION, DEPARTMENT OF COMMERCE PROCEDURES AND RULES FOR IMPLEMENTING ARTICLE 1904 OF THE NORTH AMERICAN FREE TRADE AGREEMENT Violation of a Protective Order or a...

  19. 15 CFR 908.5 - Interim reports.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... SUBMITTING REPORTS ON WEATHER MODIFICATION ACTIVITIES § 908.5 Interim reports. (a) Any person engaged in a weather modification project or activity in the United States on January 1 in any year shall submit to the... actual modification activities took place; (2) Number of days on which weather modification...

  20. Disposal facility data for the interim performance

    SciTech Connect

    Eiholzer, C.R.

    1995-05-15

    The purpose of this report is to identify and provide information on the waste package and disposal facility concepts to be used for the low-level waste tank interim performance assessment. Current concepts for the low-level waste form, canister, and the disposal facility will be used for the interim performance assessment. The concept for the waste form consists of vitrified glass cullet in a sulfur polymer cement matrix material. The waste form will be contained in a 2 {times} 2 {times} 8 meter carbon steel container. Two disposal facility concepts will be used for the interim performance assessment. These facility concepts are based on a preliminary disposal facility concept developed for estimating costs for a disposal options configuration study. These disposal concepts are based on vault type structures. None of the concepts given in this report have been approved by a Tank Waste Remediation Systems (TWRS) decision board. These concepts will only be used in th interim performance assessment. Future performance assessments will be based on approved designs.

  1. 19 CFR 207.106 - Interim measures.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 19 Customs Duties 3 2013-04-01 2013-04-01 false Interim measures. 207.106 Section 207.106 Customs Duties UNITED STATES INTERNATIONAL TRADE COMMISSION NONADJUDICATIVE INVESTIGATIONS INVESTIGATIONS OF WHETHER INJURY TO DOMESTIC INDUSTRIES RESULTS FROM IMPORTS SOLD AT LESS THAN FAIR VALUE OR FROM...

  2. 19 CFR 207.106 - Interim measures.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 19 Customs Duties 3 2012-04-01 2012-04-01 false Interim measures. 207.106 Section 207.106 Customs Duties UNITED STATES INTERNATIONAL TRADE COMMISSION NONADJUDICATIVE INVESTIGATIONS INVESTIGATIONS OF WHETHER INJURY TO DOMESTIC INDUSTRIES RESULTS FROM IMPORTS SOLD AT LESS THAN FAIR VALUE OR FROM...

  3. 19 CFR 207.106 - Interim measures.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 19 Customs Duties 3 2011-04-01 2011-04-01 false Interim measures. 207.106 Section 207.106 Customs Duties UNITED STATES INTERNATIONAL TRADE COMMISSION NONADJUDICATIVE INVESTIGATIONS INVESTIGATIONS OF WHETHER INJURY TO DOMESTIC INDUSTRIES RESULTS FROM IMPORTS SOLD AT LESS THAN FAIR VALUE OR FROM...

  4. 19 CFR 207.106 - Interim measures.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 19 Customs Duties 3 2014-04-01 2014-04-01 false Interim measures. 207.106 Section 207.106 Customs Duties UNITED STATES INTERNATIONAL TRADE COMMISSION NONADJUDICATIVE INVESTIGATIONS INVESTIGATIONS OF WHETHER INJURY TO DOMESTIC INDUSTRIES RESULTS FROM IMPORTS SOLD AT LESS THAN FAIR VALUE OR FROM...

  5. Diversified Satellite Occupations Program. Interim Report.

    ERIC Educational Resources Information Center

    Call, John Reed

    This interim report, covering the period of September 1970 to June 1971, describes a program conducted for elementary, junior high, and senior high grades. The elementary program was designed to help students develop an understanding of occupational competence. The prevention of dropouts and individualizing instruction were concerns of the junior…

  6. Interim Report - Grant AFOSR-85-0065.

    DTIC Science & Technology

    1985-06-30

    Option 065 Drafting Plotter 1 7475A 6 Pen Graphics Plotter Vendor: Imagen Corporation $ 11,266 1 Model 8/300 Laser Printer System, 512Kbytes memory 1... Imagen 8/300 Laser Printer 3 Interim Report Grant No. AFOSR-85-0065 was equivalent to the Versatec plotter for small plots and better for text output

  7. 340 waste handling facility interim safety basis

    SciTech Connect

    VAIL, T.S.

    1999-04-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

  8. Self-protection in dry recycle technologies

    SciTech Connect

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

    1995-12-01

    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.

  9. Steam drying -- Modeling and applications

    SciTech Connect

    Wimmerstedt, R.; Hager, J.

    1996-08-01

    The concept of steam drying originates from the mid of the last century. However, a broad industrial acceptance of the technique has so far not taken place. The paper deals with modelling the steam drying process and applications of steam drying within certain industrial sectors where the technique has been deemed to have special opportunities. In the modelling section the mass and heat transfer processes are described along with equilibrium, capillarity and sorption phenomena occurring in porous materials during the steam drying process. In addition existing models in the literature are presented. The applications discussed involve drying of fuels with high moisture contents, cattle feed exemplified by sugar beet pulp, lumber, paper pulp, paper and sludges. Steam drying is compared to flue gas drying of biofuels prior to combustion in a boiler. With reference to a current installation in Sweden, the exergy losses, as manifested by loss of co-generation capacity, are discussed. The energy saving potential when using steam drying of sugar beet pulp as compared to other possible plant configurations is demonstrated. Mechanical vapor recompression applied to steam drying is analyzed with reference to reported data from industrial plants. Finally, environmental advantages when using steam drying are presented.

  10. ICPP radioactive liquid and calcine waste technologies evaluation. Interim report

    SciTech Connect

    Murphy, J.A.; Pincock, L.F.; Christiansen, I.N.

    1994-06-01

    The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until recently, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, changing world events have raised questions concerning the need to recover and recycle this material. In April 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the management and disposition of radioactive wastes accumulated through reprocessing activities. Currently, 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste) and 3,800 cubic meters (m{sup 3}) of calcine waste are in inventory at the ICPP. Legal drivers and agreements exist obligating the INEL to develop, demonstrate, and implement technologies for safe and environmentally sound treatment and interim storage of radioactive liquid and calcine waste. Candidate treatment processes and waste forms are being evaluated using the Technology Evaluation and Analysis Methodology (TEAM) Model. This process allows decision makers to (1) identify optimum radioactive waste treatment and disposal form alternatives; (2) assess tradeoffs between various optimization criteria; (3) identify uncertainties in performance parameters; and (4) focus development efforts on options that best satisfy stakeholder concerns. The Systems Analysis technology evaluation presented in this document supports the DOE in selecting the most effective radioactive liquid and calcine waste management plan to implement in compliance with established regulations, court orders, and agreements.

  11. HT-9 duct cutting - IEM (Interim Examination and Maintenance) cell and mock-up testing experience at FFTF (Fast Flux Test Facility)

    SciTech Connect

    Gibbons, P.W.; Greenwell, R.K.

    1987-01-01

    This paper describes experience gained during remote cutting of the HT-9 alloy duct from an advanced fuel assembly in the Interim Examination and Maintenance (IEM) cell at the Fast Flux Test Facility (FFTF). Also described is a test program performed on mock-up equipment to develop successful cutting parameters.

  12. Hanford Single-Pass Reactor Fuel Storage Basin Demolition.

    PubMed

    Armstrong, Jason A.

    2003-02-01

    ABSTRACT The Environmental Restoration Contractor at the Hanford Site is tasked with removing auxiliary reactor structures and leaving the remaining concrete structure surrounding each reactor core. This is referred to as Interim Safe Storage. Part of placing the F Reactor into Interim Safe Storage is the demolition of the fuel storage basin, which was deactivated in 1970 by placing debris material into the basin prior to back filling with soil. Besides the debris material (wooden floor decking, handrails, and monorail pieces), the fuel storage basin contents included the possibility of spent nuclear fuel, fuel buckets, fuel spacers, process tubes, and tongs. Demolition of the fuel storage basin offered many unique radiological control challenges and innovative approaches to demolition. This paper describes how the total effective dose equivalent and contamination were controlled, how the use of a remote operated excavator was employed to remove high-dose-rate material, and how wireless technology was used to monitor changing radiological conditions.

  13. Hanford single-pass reactor fuel storage basin demolition.

    PubMed

    Armstrong, Jason A

    2003-02-01

    The Environmental Restoration Contractor at the Hanford Site is tasked with removing auxiliary reactor structures and leaving the remaining concrete structure surrounding each reactor core. This is referred to as Interim Safe Storage. Part of placing the F Reactor into Interim Safe Storage is the demolition of the fuel storage basin, which was deactivated in 1970 by placing debris material into the basin prior to back filling with soil. Besides the debris material (wooden floor decking, handrails, and monorail pieces), the fuel storage basin contents included the possibility of spent nuclear fuel, fuel buckets, fuel spacers, process tubes, and tongs. Demolition of the fuel storage basin offered many unique radiological control challenges and innovative approaches to demolition. This paper describes how the total effective dose equivalent and contamination were controlled, how the use of a remote operated excavator was employed to remove high-dose-rate material, and how wireless technology was used to monitor changing radiological conditions.

  14. century drying

    NASA Astrophysics Data System (ADS)

    Cook, Benjamin I.; Smerdon, Jason E.; Seager, Richard; Coats, Sloan

    2014-11-01

    Global warming is expected to increase the frequency and intensity of droughts in the twenty-first century, but the relative contributions from changes in moisture supply (precipitation) versus evaporative demand (potential evapotranspiration; PET) have not been comprehensively assessed. Using output from a suite of general circulation model (GCM) simulations from phase 5 of the Coupled Model Intercomparison Project, projected twenty-first century drying and wetting trends are investigated using two offline indices of surface moisture balance: the Palmer Drought Severity Index (PDSI) and the Standardized Precipitation Evapotranspiration Index (SPEI). PDSI and SPEI projections using precipitation and Penman-Monteith based PET changes from the GCMs generally agree, showing robust cross-model drying in western North America, Central America, the Mediterranean, southern Africa, and the Amazon and robust wetting occurring in the Northern Hemisphere high latitudes and east Africa (PDSI only). The SPEI is more sensitive to PET changes than the PDSI, especially in arid regions such as the Sahara and Middle East. Regional drying and wetting patterns largely mirror the spatially heterogeneous response of precipitation in the models, although drying in the PDSI and SPEI calculations extends beyond the regions of reduced precipitation. This expansion of drying areas is attributed to globally widespread increases in PET, caused by increases in surface net radiation and the vapor pressure deficit. Increased PET not only intensifies drying in areas where precipitation is already reduced, it also drives areas into drought that would otherwise experience little drying or even wetting from precipitation trends alone. This PET amplification effect is largest in the Northern Hemisphere mid-latitudes, and is especially pronounced in western North America, Europe, and southeast China. Compared to PDSI projections using precipitation changes only, the projections incorporating both

  15. Remote weighing of irradiated fuel pins at FFTF

    SciTech Connect

    Anglesey, M.O.; Romrell, D.M.

    1986-01-01

    This paper describes the testing and operations of a remotely operated fuel pin weighing system developed to identify fuel pins with breached cladding in the interim examination and maintenance (IEM) cell at the Fast Flux Test Facility (FFTF) located near Richland, Washington. The IEM cell is a vertical hot cell located within the FFTF containment building that was designed for disassembly and reassembly of experiments and fuel assemblies.

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

    NASA Astrophysics Data System (ADS)

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

    2015-03-01

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

  17. Drying Milk With Boiler Exhaust

    NASA Technical Reports Server (NTRS)

    Broussard, M. R.

    1984-01-01

    Considerable energy saved in powdered-milk industry. Only special requirement boiler fired with natural gas or other clean fuel. Boiler flue gas fed to spray drier where it directly contacts product to be dried. Additional heat supplied by auxillary combustor when boiler output is low. Approach adaptable to existing plants with minimal investment because most already equipped with natural-gas-fired boilers.

  18. Lessons learned from the Siting Process of an Interim Storage Facility in Spain - 12024

    SciTech Connect

    Lamolla, Meritxell Martell

    2012-07-01

    On 29 December 2009, the Spanish government launched a site selection process to host a centralised interim storage facility for spent fuel and high-level radioactive waste. It was an unprecedented call for voluntarism among Spanish municipalities to site a controversial facility. Two nuclear municipalities, amongst a total of thirteen municipalities from five different regions, presented their candidatures to host the facility in their territories. For two years the government did not make a decision. Only in November 30, 2011, the new government elected on 20 November 2011 officially selected a non-nuclear municipality, Villar de Canas, for hosting this facility. This paper focuses on analysing the factors facilitating and hindering the siting of controversial facilities, in particular the interim storage facility in Spain. It demonstrates that involving all stakeholders in the decision-making process should not be underestimated. In the case of Spain, all regional governments where there were candidate municipalities willing to host the centralised interim storage facility, publicly opposed to the siting of the facility. (author)

  19. Dry cell battery poisoning

    MedlinePlus

    Batteries - dry cell ... Acidic dry cell batteries contain: Manganese dioxide Ammonium chloride Alkaline dry cell batteries contain: Sodium hydroxide Potassium hydroxide Lithium dioxide dry cell batteries ...

  20. 46 CFR 108.489 - Helicopter fueling facilities.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... containment systems around marine portable tanks, fuel transfer pumps and fuel hose reels: (1) Protein foam at...) 22.5 kilograms (50 pounds) of dry chemical (B-V semi-portable) for each fueling facility of up to...

  1. 78 FR 67442 - Congestion Mitigation and Air Quality Improvement Program Interim Guidance

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-12

    ... Federal Highway Administration Congestion Mitigation and Air Quality Improvement Program Interim Guidance... Mitigation and Air Quality Improvement (CMAQ) Program (Interim Guidance). The Interim Guidance revises CMAQ....gov/environment/air_quality/cmaq/policy_and_guidance/2008_guidance/ guidance/. DATES: This...

  2. 76 FR 15028 - Airport Improvement Program (AIP): Interim Policy Regarding Access to Airports From Residential...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-18

    ... Federal Aviation Administration Airport Improvement Program (AIP): Interim Policy Regarding Access to Airports From Residential Property AGENCY: Federal Aviation Administration (FAA). ACTION: Interim policy; amendment to sponsor grant assurance 5. SUMMARY: This action adopts an interim policy amending...

  3. Biocides for the Battlefield - Interim Report

    DTIC Science & Technology

    2010-09-24

    for public release; distribution is unlimited. James H . Wynne Materials Chemistry Branch Chemistry Division Preston a. Fulmer Chemical Dynamics and...b. ABSTRACT c. THIS PAGE 18. NUMBER OF PAGES 17. LIMITATION OF ABSTRACT Biocides for the Battlefield—Interim Report James H . Wynne and Preston A...distribution is unlimited. Unclassified Unclassified Unclassified SAR 25 James H . Wynne (202) 404-4010 The ability to disinfect surfaces and manufacture self

  4. Cold vacuum drying facility 90% design review

    SciTech Connect

    O`Neill, C.T.

    1997-05-02

    This document contains review comment records for the CVDF 90% design review. Spent fuels retrieved from the K Basins will be dried at the CVDF. It has also been recommended that the Multi-Conister Overpacks be welded, inspected, and repaired at the CVD Facility before transport to dry storage.

  5. Compilation of interim technical research memoranda. Volume I

    SciTech Connect

    Shanahan, W.R.

    1984-04-01

    Four interim technical research memoranda are presented that describe the results of numerical simulations designed to investigate the dynamics of energetic plasma beams propagating across magnetic fields.

  6. Fuel performance in water storage

    SciTech Connect

    Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

    1993-11-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL) for the Department of Energy (DOE). A variety of different types of fuels have been stored there since the 1950`s prior to reprocessing for uranium recovery. In April of 1992, the DOE decided to end fuel reprocessing, changing the mission at ICPP. Fuel integrity in storage is now viewed as long term until final disposition is defined and implemented. Thus, the condition of fuel and storage equipment is being closely monitored and evaluated to ensure continued safe storage. There are four main areas of fuel storage at ICPP: an original underwater storage facility (CPP-603), a modern underwater storage facility (CPP-666), and two dry fuel storage facilities. The fuels in storage are from the US Navy, DOE (and its predecessors the Energy Research and Development Administration and the Atomic Energy Commission), and other research programs. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels. In the underwater storage basins, fuels are clad with stainless steel, zirconium, and aluminum. Also included in the basin inventory is canned scrap material. The dry fuel storage contains primarily graphite and aluminum type fuels. A total of 55 different fuel types are currently stored at the Idaho Chemical Processing Plant. The corrosion resistance of the barrier material is of primary concern in evaluating the integrity of the fuel in long term water storage. The barrier material is either the fuel cladding (if not canned) or the can material.

  7. Colorful drying.

    PubMed

    Lakio, Satu; Heinämäki, Jyrki; Yliruusi, Jouko

    2010-03-01

    Drying is one of the standard unit operations in the pharmaceutical industry and it is important to become aware of the circumstances that dominate during the process. The purpose of this study was to test microcapsulated thermochromic pigments as heat indicators in a fluid bed drying process. The indicator powders were manually granulated with alpha-lactose monohydrate resulting in three particle-size groups. Also, pellets were coated with the indicator powders. The granules and pellets were fluidized in fluid bed dryer to observe the progress of the heat flow in the material and to study the heat indicator properties of the indicator materials. A tristimulus colorimeter was used to measure CIELAB color values. Color indicator for heat detection can be utilized to test if the heat-sensitive API would go through physical changes during the pharmaceutical drying process. Both the prepared granules and pellets can be used as heat indicator in fluid bed drying process. The colored heat indicators give an opportunity to learn new aspects of the process at real time and could be exploded, for example, for scaling-up studies.

  8. Use of a sub-gasket and soft gas diffusion layer to mitigate mechanical degradation of a hydrocarbon membrane for polymer electrolyte fuel cells in wet-dry cycling

    NASA Astrophysics Data System (ADS)

    Ishikawa, Hiroshi; Teramoto, Takeshi; Ueyama, Yasuhiro; Sugawara, Yasushi; Sakiyama, Yoko; Kusakabe, Masato; Miyatake, Kenji; Uchida, Makoto

    2016-09-01

    The mechanical durability of hydrocarbon (HC) membranes, used for polymer electrolyte fuel cells (PEFCs), was evaluated by the United States Department of Energy (USDOE) stress protocol involving wet-dry cycling, and the degradation mechanism is discussed. The HC membrane ruptured in the edge region of the membrane electrode assembly (MEA) after 300 cycles due to a concentration of the mechanical stress. Post-test analysis of stress-strain measurements revealed that the membrane mechanical strain decreased more than 80% in the edge region of the MEA and about 50% in the electrode region, compared with the pristine condition. Size exclusion chromatography (SEC) indicated that the average molecular weight of the HC polymer increased slightly, indicating some cross-linking, while the IEC decreased slightly, indicating ionomer degradation. As a result of two types of modifications, a sub-gasket (SG) and a soft gas diffusion layer (GDL) in the MEA edge region, the mechanical stress decreased, and the durability increased, the membrane lasting more than 30,000 cycles without mechanical failure.

  9. Review of progress in the Canadian nuclear fuel waste management program

    SciTech Connect

    Lyon, R.B.; Johnson, L.H.

    1986-01-01

    The Canadian Nuclear Fuel Waste Management Program is reviewed, illustrating the progress that has been made in assessing the concept of disposal of nuclear fuel waste in plutonic rock of the Canadian Shield. Research is being conducted into used fuel storage and transportation, fuel waste immobilization, site characterization and selection methods, and performance assessment modelling. Details of achievements in these areas are outlined, and results of the most recent interim assessment are discussed.

  10. 47 CFR 73.404 - Interim hybrid IBOC DAB operation.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... RADIO BROADCAST SERVICES Digital Audio Broadcasting § 73.404 Interim hybrid IBOC DAB operation. (a) The... test operation pursuant to § 73.1620, may commence interim hybrid IBOC DAB operation with digital... Report and Order in MM Docket No. 99-325, as revised in the Media Bureau's subsequent Order in MM...

  11. 50 CFR 660.720 - Interim protection for sea turtles.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 50 Wildlife and Fisheries 11 2011-10-01 2011-10-01 false Interim protection for sea turtles. 660.720 Section 660.720 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND... Migratory Fisheries § 660.720 Interim protection for sea turtles. (a) Until the effective date of §§...

  12. 50 CFR 660.720 - Interim protection for sea turtles.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 50 Wildlife and Fisheries 13 2013-10-01 2013-10-01 false Interim protection for sea turtles. 660.720 Section 660.720 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND... Migratory Fisheries § 660.720 Interim protection for sea turtles. (a) Until the effective date of §§...

  13. 50 CFR 660.720 - Interim protection for sea turtles.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Interim protection for sea turtles. 660.720 Section 660.720 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND... Migratory Fisheries § 660.720 Interim protection for sea turtles. (a) Until the effective date of §§...

  14. 50 CFR 660.720 - Interim protection for sea turtles.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 50 Wildlife and Fisheries 13 2014-10-01 2014-10-01 false Interim protection for sea turtles. 660.720 Section 660.720 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND... Migratory Fisheries § 660.720 Interim protection for sea turtles. (a) Until the effective date of §§...

  15. 50 CFR 660.720 - Interim protection for sea turtles.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 50 Wildlife and Fisheries 13 2012-10-01 2012-10-01 false Interim protection for sea turtles. 660.720 Section 660.720 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND... Migratory Fisheries § 660.720 Interim protection for sea turtles. (a) Until the effective date of §§...

  16. 46 CFR 308.203 - Amount insured under interim binder.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 8 2010-10-01 2010-10-01 false Amount insured under interim binder. 308.203 Section 308.203 Shipping MARITIME ADMINISTRATION, DEPARTMENT OF TRANSPORTATION EMERGENCY OPERATIONS WAR RISK INSURANCE War Risk Protection and Indemnity Insurance § 308.203 Amount insured under interim binder....

  17. 45 CFR 1606.13 - Interim and termination funding; reprogramming.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 45 Public Welfare 4 2010-10-01 2010-10-01 false Interim and termination funding; reprogramming. 1606.13 Section 1606.13 Public Welfare Regulations Relating to Public Welfare (Continued) LEGAL SERVICES CORPORATION TERMINATION AND DEBARMENT PROCEDURES; RECOMPETITION § 1606.13 Interim and...

  18. Presidential Transition: The Experience of Two Community College Interim Presidents

    ERIC Educational Resources Information Center

    Thompson, Matthew D.

    2010-01-01

    The purpose of this qualitative case study was to understand the experiences of two community college interim presidents, their characteristics, and how they led institutions following an abrupt presidential departure. There were two fundamental questions framing this research study, 1. How do two interim community college presidents lead…

  19. 16 CFR 1203.53 - Interim safety standards.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Interim safety standards. 1203.53 Section 1203.53 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT SAFETY ACT REGULATIONS..., Through March 10, 1999 § 1203.53 Interim safety standards. (a) Bicycle helmets must comply with one...

  20. 46 CFR 308.303 - Amounts insured under interim binder.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 8 2010-10-01 2010-10-01 false Amounts insured under interim binder. 308.303 Section 308.303 Shipping MARITIME ADMINISTRATION, DEPARTMENT OF TRANSPORTATION EMERGENCY OPERATIONS WAR RISK INSURANCE Second Seamen's War Risk Insurance § 308.303 Amounts insured under interim binder. The...

  1. Optimal timing for interim analyses in clinical trials.

    PubMed

    Togo, Kanae; Iwasaki, Manabu

    2013-01-01

    In clinical trials, interim analyses are often performed before the completion of the trial. The intention is to possibly terminate the trial early or adjust the sample size. The time of conducting an interim analysis affects the probability of the early termination and the number of subjects enrolled until the interim analysis. This influences the expected total number of subjects. In this study, we examine the optimal time for conducting interim analyses with a view to minimizing the expected total sample size. It is found that regardless of the effect size, the optimal time of one interim analysis for the early termination is approximately two-thirds of the planned observations for the O'Brien-Fleming type of spending function and approximately half of the planned observations for the Pocock type when the subject enrollment is halted for the interim analysis. When the subject enrollment is continuous throughout the trial, the optimal time for the interim analysis varies according to the follow-up duration. We also consider the time for one interim analysis including the sample size adjustment in terms of minimizing the expected total sample size.

  2. Staff Reactions to Interim Leadership in a Student Affairs Organization

    ERIC Educational Resources Information Center

    Jones, Robin D.

    2011-01-01

    Interim leadership appointments in higher education are a common strategy used to fill leadership gaps in executive positions. Because student affairs executives are particularly vulnerable to high turnover rates, interim appointments are becoming more widespread. Even with the prevalence of this trend, little attention has been given to the…

  3. 7 CFR 280.1 - Interim disaster procedures.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 4 2010-01-01 2010-01-01 false Interim disaster procedures. 280.1 Section 280.1 Agriculture Regulations of the Department of Agriculture (Continued) FOOD AND NUTRITION SERVICE, DEPARTMENT OF....1 Interim disaster procedures. The Secretary shall, after consultation with the official...

  4. 40 CFR 281.23 - Schedule for interim approval.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Schedule for interim approval. 281.23 Section 281.23 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES... Schedule for interim approval. For a state program that must modify its statutory or...

  5. 47 CFR 51.715 - Interim transport and termination pricing.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 47 Telecommunication 3 2010-10-01 2010-10-01 false Interim transport and termination pricing. 51... Telecommunications Traffic § 51.715 Interim transport and termination pricing. (a) Upon request from a... shall provide transport and termination of telecommunications traffic immediately under an...

  6. Presidential Transition: The Experience of Two Community College Interim Presidents

    ERIC Educational Resources Information Center

    Thompson, Matthew; Cooper, Robyn; Ebbers, Larry

    2012-01-01

    The purpose of this case study was to understand the experiences of two community college interim presidents, their characteristics, and how they led institutions following an abrupt presidential departure. There were two fundamental questions framing this study: 1. How do two interim community college presidents lead community colleges during a…

  7. The Predictive and Instructional Value of Interim Assessments

    ERIC Educational Resources Information Center

    Pon, Kathleen

    2013-01-01

    This mixed design study investigated the predictive and instructional uses of two different types of interim mathematics assessments given in two different districts. One district administered the same summative type of assessment three times a year, while the other district administered a different interim assessment after six-week intervals of…

  8. 40 CFR 270.64 - Interim permits for UIC wells.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Interim permits for UIC wells. 270.64 Section 270.64 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES... § 270.64 Interim permits for UIC wells. The Director may issue a permit under this part to any Class...

  9. 46 CFR 308.103 - Insured amounts under interim binder.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 8 2010-10-01 2010-10-01 false Insured amounts under interim binder. 308.103 Section 308.103 Shipping MARITIME ADMINISTRATION, DEPARTMENT OF TRANSPORTATION EMERGENCY OPERATIONS WAR RISK INSURANCE War Risk Hull and Disbursements Insurance § 308.103 Insured amounts under interim binder....

  10. 42 CFR 418.307 - Periodic interim payments.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 42 Public Health 3 2010-10-01 2010-10-01 false Periodic interim payments. 418.307 Section 418.307... (CONTINUED) MEDICARE PROGRAM HOSPICE CARE Payment for Hospice Care § 418.307 Periodic interim payments. Subject to the provisions of § 413.64(h) of this chapter, a hospice may elect to receive periodic...

  11. 40 CFR 155.56 - Interim registration review decision.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Interim registration review decision... PROGRAMS REGISTRATION STANDARDS AND REGISTRATION REVIEW Registration Review Procedures § 155.56 Interim registration review decision. The Agency may issue, when it determines it to be appropriate, an...

  12. 39 CFR 211.4 - Interim personnel regulations.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 39 Postal Service 1 2010-07-01 2010-07-01 false Interim personnel regulations. 211.4 Section 211.4 Postal Service UNITED STATES POSTAL SERVICE ORGANIZATION AND ADMINISTRATION APPLICATION OF REGULATIONS § 211.4 Interim personnel regulations. (a) Continuation of Personnel Regulations of the Post...

  13. Transuranic waste storage and assay facility (TRUSAF) interim safety basis

    SciTech Connect

    Gibson, K.D.

    1995-09-01

    The TRUSAF ISB is based upon current facility configuration and procedures. The purpose of the document is to provide the basis for interim operation or restrictions on interim operations and the authorization basis for the TRUSAF at the Hanford Site. The previous safety analysis document TRUSAF hazards Identification and Evaluation (WHC 1977) is superseded by this document.

  14. Statistical Profile of Children and Mothers in Afghanistan. Interim Edition.

    ERIC Educational Resources Information Center

    United Nations Children's Fund, Kabul (Afghanistan).

    This interim report is an updating of the 1977 Statistical Profile of Children and Mothers in Afghanistan. The interim report reflects the significant changes in policies brought about by the Saur Revolution establishing the Democratic Republic of Afghanistan in 1978. A comprehensive revision of the report is expected when the new government's…

  15. 28 CFR 94.41 - Interim emergency payment.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 28 Judicial Administration 2 2012-07-01 2012-07-01 false Interim emergency payment. 94.41 Section 94.41 Judicial Administration DEPARTMENT OF JUSTICE (CONTINUED) CRIME VICTIM SERVICES International Terrorism Victim Expense Reimbursement Program Payment of Claims § 94.41 Interim emergency...

  16. 28 CFR 94.41 - Interim emergency payment.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 28 Judicial Administration 2 2014-07-01 2014-07-01 false Interim emergency payment. 94.41 Section 94.41 Judicial Administration DEPARTMENT OF JUSTICE (CONTINUED) CRIME VICTIM SERVICES International Terrorism Victim Expense Reimbursement Program Payment of Claims § 94.41 Interim emergency...

  17. 28 CFR 94.41 - Interim emergency payment.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 28 Judicial Administration 2 2013-07-01 2013-07-01 false Interim emergency payment. 94.41 Section 94.41 Judicial Administration DEPARTMENT OF JUSTICE (CONTINUED) CRIME VICTIM SERVICES International Terrorism Victim Expense Reimbursement Program Payment of Claims § 94.41 Interim emergency...

  18. 28 CFR 94.41 - Interim emergency payment.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 28 Judicial Administration 2 2011-07-01 2011-07-01 false Interim emergency payment. 94.41 Section 94.41 Judicial Administration DEPARTMENT OF JUSTICE (CONTINUED) CRIME VICTIM SERVICES International Terrorism Victim Expense Reimbursement Program Payment of Claims § 94.41 Interim emergency...

  19. 49 CFR 37.193 - Interim service requirements.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... WITH DISABILITIES (ADA) Over-the-Road Buses (OTRBs) § 37.193 Interim service requirements. (a) Until... accessible bus, the operator has met this requirement. (2) Interim service under this paragraph (a) is not... percent of which consists of accessible buses (e.g., a small fixed-route operator who exclusively...

  20. 49 CFR 37.193 - Interim service requirements.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... WITH DISABILITIES (ADA) Over-the-Road Buses (OTRBs) § 37.193 Interim service requirements. (a) Until... accessible bus, the operator has met this requirement. (2) Interim service under this paragraph (a) is not... percent of which consists of accessible buses (e.g., a small fixed-route operator who exclusively...

  1. 49 CFR 37.193 - Interim service requirements.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... WITH DISABILITIES (ADA) Over-the-Road Buses (OTRBs) § 37.193 Interim service requirements. (a) Until... accessible bus, the operator has met this requirement. (2) Interim service under this paragraph (a) is not... percent of which consists of accessible buses (e.g., a small fixed-route operator who exclusively...

  2. 49 CFR 106.35 - Interim final rule.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY... requirements and their effective date. PHMSA may issue an interim final rule if it finds, for good cause, that... comments, as well as any other relevant documents, PHMSA may revise the interim final rule and then issue...

  3. 49 CFR 106.35 - Interim final rule.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY... requirements and their effective date. PHMSA may issue an interim final rule if it finds, for good cause, that... comments, as well as any other relevant documents, PHMSA may revise the interim final rule and then issue...

  4. 49 CFR 106.35 - Interim final rule.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY... requirements and their effective date. PHMSA may issue an interim final rule if it finds, for good cause, that... comments, as well as any other relevant documents, PHMSA may revise the interim final rule and then issue...

  5. 49 CFR 106.35 - Interim final rule.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY... requirements and their effective date. PHMSA may issue an interim final rule if it finds, for good cause, that... comments, as well as any other relevant documents, PHMSA may revise the interim final rule and then issue...

  6. Augmented Reality Mentor for Training Maintenance Procedures: Interim Assessment

    DTIC Science & Technology

    2014-08-01

    ARI Research Note 2014-04 Augmented Reality Mentor for Training Maintenance Procedures: Interim Assessment Louise...2014 4. TITLE AND SUBTITLE Augmented Reality Mentor for Training Maintenance Procedures: Interim Assessment 5a. CONTRACT OR GRANT NUMBER...Representative and Subject Matter POC: Dr. William R. Bickley 14. ABSTRACT (Maximum 200 words): The Augmented Reality Mentor is a 2-yr advanced

  7. 10 CFR 590.403 - Emergency interim orders.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... DEPARTMENT OF ENERGY (CONTINUED) NATURAL GAS (ECONOMIC REGULATORY ADMINISTRATION) ADMINISTRATIVE PROCEDURES WITH RESPECT TO THE IMPORT AND EXPORT OF NATURAL GAS Opinions and Orders § 590.403 Emergency interim... and issue an emergency interim order authorizing the import or export of natural gas. After...

  8. 10 CFR 590.403 - Emergency interim orders.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... DEPARTMENT OF ENERGY (CONTINUED) NATURAL GAS (ECONOMIC REGULATORY ADMINISTRATION) ADMINISTRATIVE PROCEDURES WITH RESPECT TO THE IMPORT AND EXPORT OF NATURAL GAS Opinions and Orders § 590.403 Emergency interim... and issue an emergency interim order authorizing the import or export of natural gas. After...

  9. 42 CFR 418.307 - Periodic interim payments.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... (CONTINUED) MEDICARE PROGRAM HOSPICE CARE Payment for Hospice Care § 418.307 Periodic interim payments. Subject to the provisions of § 413.64(h) of this chapter, a hospice may elect to receive periodic interim... PIP method unless the hospice requests a longer fixed interval (not to exceed one month)...

  10. 47 CFR 73.404 - Interim hybrid IBOC DAB operation.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... RADIO BROADCAST SERVICES Digital Audio Broadcasting § 73.404 Interim hybrid IBOC DAB operation. (a) The... test operation pursuant to § 73.1620, may commence interim hybrid IBOC DAB operation with digital... No. 99-325. FM stations are permitted to operate with hybrid digital effective radiated power...

  11. 47 CFR 73.404 - Interim hybrid IBOC DAB operation.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... RADIO BROADCAST SERVICES Digital Audio Broadcasting § 73.404 Interim hybrid IBOC DAB operation. (a) The... test operation pursuant to § 73.1620, may commence interim hybrid IBOC DAB operation with digital... No. 99-325. FM stations are permitted to operate with hybrid digital effective radiated power...

  12. 17 CFR 210.8-03 - Interim financial statements.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... filed with the interim financial statements. Interim financial statements shall include a balance sheet as of the end of the issuer's most recent fiscal quarter, a balance sheet as of the end of the... date of such balance sheet and the comparable period of the preceding fiscal year. (a) Condensed...

  13. Central waste complex interim safety basis

    SciTech Connect

    Cain, F.G.

    1995-05-15

    This interim safety basis provides the necessary information to conclude that hazards at the Central Waste Complex are controlled and that current and planned activities at the CWC can be conducted safely. CWC is a multi-facility complex within the Solid Waste Management Complex that receives and stores most of the solid wastes generated and received at the Hanford Site. The solid wastes that will be handled at CWC include both currently stored and newly generated low-level waste, low-level mixed waste, contact-handled transuranic, and contact-handled TRU mixed waste.

  14. 50 CFR 259.30 - Application for Interim Capital Construction Fund Agreement (“Interim CCF Agreement”).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 50 Wildlife and Fisheries 7 2010-10-01 2010-10-01 false Application for Interim Capital Construction Fund Agreement (âInterim CCF Agreementâ). 259.30 Section 259.30 Wildlife and Fisheries NATIONAL... FISHERIES CAPITAL CONSTRUCTION FUND Capital Construction Fund Agreement § 259.30 Application for...

  15. 76 FR 53813 - Dried Prunes Produced in California; Decreased Assessment Rate

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-08-30

    ... Agricultural Marketing Service 7 CFR Part 993 Dried Prunes Produced in California; Decreased Assessment Rate AGENCY: Agricultural Marketing Service, USDA. ACTION: Interim rule with request for comments. SUMMARY... order is effective under the Agricultural Marketing Agreement Act of 1937, as amended (7 U.S.C....

  16. 78 FR 19148 - Shielding and Radiation Protection Review Effort and Licensing Conditions for Dry Storage...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-29

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 Shielding and Radiation Protection Review Effort and Licensing Conditions for Dry Storage Applications AGENCY: Nuclear Regulatory Commission. ACTION: Draft interim staff guidance; request for public comment. SUMMARY: The U.S. Nuclear Regulatory Commission...

  17. Disposability Assessment: Aluminum-Based Spent Nuclear Fuel Forms

    SciTech Connect

    Vinson, D.W.

    1998-11-06

    This report provides a technical assessment of the Melt-Dilute and Direct Al-SNF forms in disposable canisters with respect to meeting the requirements for disposal in the Mined Geologic Disposal System (MGDS) and for interim dry storage in the Treatment and Storage Facility (TSF) at SRS.

  18. Generator Set Durability Testing Using 25% ATJ Fuel Blend

    DTIC Science & Technology

    2016-02-01

    FMTV – Family of Medium Tactical Vehicles GEP – General Engine Products HC – hydrocarbon HEUI – hydraulically actuated, electronically controlled...loggers: Campbell Scientific model CR3000. Thirty thermocouples were used on each generator , along with five pressure transducers. Voltage, current...UNCLASSIFIED UNCLASSIFIED GENERATOR SET DURABILITY TESTING USING 25% ATJ FUEL BLEND INTERIM REPORT TFLRF No. 476 by Gregory A. T

  19. Interim treatment: Bridging delays to opioid treatment access

    PubMed Central

    Sigmon, Stacey C.

    2015-01-01

    Objective Despite the undisputed effectiveness of agonist maintenance for opioid dependence, individuals can remain on waitlists for months, during which they are at significant risk for morbidity and mortality. To mitigate these risks, the Food and Drug Administration in 1993 approved interim treatment, involving daily medication + emergency counseling only, when only a waitlist is otherwise available. We review the published research in the 20 years since the approval of interim opioid treatment. Methods A literature search was conducted to identify all randomized trials evaluating the efficacy of interim treatment for opioid-dependent patients awaiting comprehensive treatment. Results Interim opioid treatment has been evaluated in four controlled trials to date. In three, interim treatment was compared to waitlist or placebo control conditions and produced greater outcomes on measures of illicit opioid use, retention, criminality and likelihood of entry into comprehensive treatment. In the fourth, interim treatment was compared to standard methadone maintenance and produced comparable outcomes in illicit opioid use, retention and criminal activity. Conclusions Interim treatment significantly reduces patient and societal risks when conventional treatment is unavailable. Further research is needed to examine the generality of these findings, further enhance outcomes, and identify the patient characteristics which predict treatment response. PMID:25937593

  20. Design of dry barriers for containment of contaminants in unsaturated soils

    SciTech Connect

    Morris, C.E.; Thomson, B.M.; Stormont, J.C.

    1997-12-31

    A dry barrier is a region of very dry conditions in unsaturated soil that prevents vertical migration of water created by circulating dry air through the formation. Dry soil creates a barrier to vertical water movement by decreasing the soil`s hydraulic conductivity, a concept also used in capillary barriers. A dry barrier may be a viable method for providing containment of a contaminant plume in a setting with a thick unsaturated zone and dry climate. The principal factors which determine the feasibility of a dry barrier include: (1) an and environment, (2) thick vadose zone, and (3) the ability to circulate air through the vadose zone. This study investigated the technical and economic considerations associated with creating a dry barrier to provide containment of a hypothetical 1 ha aqueous contaminant plume. The concept appears to be competitive with other interim containment methods such as ground freezing.

  1. Fuel assembly cooling experience at the FFTF IEM cell

    SciTech Connect

    McGuinness, P.W.

    1985-11-01

    To date, 13 fuel assemblies requiring forced cooling have been processed through the Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell. Of these, two assemblies experienced overtemperature conditions due to inadequate forced cooling. Both of the occurrences have contributed significantly to the process of learning how to operate a fuel assembly cooling system remotely in an argon atmosphere hot cell. Many innovations have been made to the cooling system to enhance safety and increase productivity, and are briefly described.

  2. Fuel assembly cooling experience at the FFTF/IEM cell

    SciTech Connect

    McGuinness, P.W.

    1985-01-01

    In the Fast Flux Test Facility (FFTF), sodium wetted irradiated fuel assemblies are discharged to the Interim Examination and Maintenance (IEM) Cell for disassembly and post-irradiation examination in an inert argon atmosphere. While in the IEM Cell, fuel assemblies are cooled by the IEM Cell Subassembly Cooling System. This paper describes the cooling system design, performance, and lessons learned, including a discussion of two overtemperature incidents. 2 refs., 6 figs.

  3. Evaluation of MHTGR fuel reliability

    SciTech Connect

    Wichner, R.P.; Barthold, W.P.

    1992-07-01

    Modular High-Temperature Gas-Cooled Reactor (MHTGR) concepts that house the reactor vessel in a tight but unsealed reactor building place heightened importance on the reliability of the fuel particle coatings as fission product barriers. Though accident consequence analyses continue to show favorable results, the increased dependence on one type of barrier, in addition to a number of other factors, has caused the Nuclear Regulatory Commission (NRC) to consider conservative assumptions regarding fuel behavior. For this purpose, the concept termed ``weak fuel`` has been proposed on an interim basis. ``Weak fuel`` is a penalty imposed on consequence analyses whereby the fuel is assumed to respond less favorably to environmental conditions than predicted by behavioral models. The rationale for adopting this penalty, as well as conditions that would permit its reduction or elimination, are examined in this report. The evaluation includes an examination of possible fuel-manufacturing defects, quality-control procedures for defect detection, and the mechanisms by which fuel defects may lead to failure.

  4. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  5. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  6. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  7. Spent-fuel-storage alternatives

    SciTech Connect

    Not Available

    1980-01-01

    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)

  8. Realistic Challenges to Optimize the Closed Nuclear Fuel Cycle

    SciTech Connect

    Shigeo Nomura; Hisao Ojima; Hirohisa Ishikawa

    2002-07-01

    The first generation nuclear fuel cycle system will be closed in Japan by 2010 with the successful combination of LWR-MOX, industrial operation of reprocessing and MOX fuel fabrication plants, interim storage of spent fuel, and plans for the location of final disposal of radioactive wastes. To tailor nuclear power as the globally sustainable and environmentally benign energy source for future society, however, an advanced fuel cycle option is needed that is well suited to the fast reactor. It should recycle fissile materials of plutonium and uranium whilst transmuting minor actinides and generating minimum waste. (authors)

  9. PROJECT W-551 INTERIM PRETREATMENT SYSTEM PRECONCEPTUAL CANDIDATE TECHNOLOGY DESCRIPTIONS

    SciTech Connect

    MAY TH

    2008-08-12

    The Office of River Protection (ORP) has authorized a study to recommend and select options for interim pretreatment of tank waste and support Waste Treatment Plant (WTP) low activity waste (LAW) operations prior to startup of all the WTP facilities. The Interim Pretreatment System (IPS) is to be a moderately sized system which separates entrained solids and 137Cs from tank waste for an interim time period while WTP high level waste vitrification and pretreatment facilities are completed. This study's objective is to prepare pre-conceptual technology descriptions that expand the technical detail for selected solid and cesium separation technologies. This revision includes information on additional feed tanks.

  10. Retention of long-term interim restorations with sodium fluoride enriched interim cement

    NASA Astrophysics Data System (ADS)

    Strash, Carolyn

    Purpose: Interim fixed dental prostheses, or "provisional restorations", are fabricated to restore teeth when definitive prostheses are made indirectly. Patients undergoing extensive prosthodontic treatment frequently require provisionalization for several months or years. The ideal interim cement would retain the restoration for as long as needed and still allow for ease of removal. It would also avoid recurrent caries by preventing demineralization of tooth structure. This study aims to determine if adding sodium fluoride varnish to interim cement may assist in the retention of interim restorations. Materials and methods: stainless steel dies representing a crown preparation were fabricated. Provisional crowns were milled for the dies using CAD/CAM technology. Crowns were provisionally cemented onto the dies using TempBond NE and NexTemp provisional cements as well as a mixture of TempBond NE and Duraphat fluoride varnish. Samples were stored for 24h then tested or thermocycled for 2500 or 5000 cycles before being tested. Retentive strength of each cement was recorded using a universal testing machine. Results: TempBond NE and NexTemp cements performed similarly when tested after 24h. The addition of Duraphat significantly decreased the retention when added to TempBond NE. NexTemp cement had high variability in retention over all tested time periods. Thermocycling for 2500 and 5000 cycles significantly decreased the retention of all cements. Conclusions: The addition of Duraphat fluoride varnish significantly decreased the retention of TempBond NE and is therefore not recommended for clinical use. Thermocycling significantly reduced the retention of TempBond NE and NexTemp. This may suggest that use of these cements for three months, as simulated in this study, is not recommended.

  11. Characterization of In-Drum Drying Products

    SciTech Connect

    Kroselj, V.; Jankovic, M.; Skanata, D.; Medakovic, S.; Harapin, D.; Hertl, B.

    2006-07-01

    A few years ago Krsko NPP decided to introduce In-Drum Drying technology for treatment and conditioning of evaporator concentrates and spent ion resins. The main reason to employ this technology was the need for waste volume reduction and experience with vermiculite-cement solidification that proved inadequate for Krsko NPP. Use of In-Drum Drying technology was encouraged by good experience in the field at some German and Spanish NPP's. In the paper, solidification techniques in vermiculite-cement matrix and In-Drum Drying System are described briefly. The resulting waste forms (so called solidification and dryer products) and containers that are used for interim storage of these wastes are described as well. A comparison of the drying versus solidification technology is performed and advantages as well as disadvantages are underlined. Experience gained during seven years of system operation has shown that crying technology resulted in volume reduction by factor of 20 for evaporator concentrates, and by factor of 5 for spent ion resin. Special consideration is paid to the characterization of dryer products. For evaporator concentrates the resulting waste form is a solid salt block with up to 5% bound water. It is packaged in stainless steel drums (net volume of 200 l) with bolted lids and lifting rings. The fluidized spent ion resins (primary and blow-down) are sluiced into the spent resin drying tank. The resin is dewatered and dried by electrical jacket heaters. The resulting waste (i.e. fine granulates) is directly discharged into a shielded stainless steel drum with bolted lid and lifting rings. Characterization of both waste forms has been performed in accordance with recommendations given in Characterization of Radioactive Waste Forms and Packages issued by International Atomic Energy Agency, 1997. This means that radiological, chemical, physical, mechanical, biological and thermal properties of the waste form has been taken into consideration. In the paper

  12. 40 CFR 600.117 - Interim provisions.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... ECONOMY AND GREENHOUSE GAS EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy and Carbon-Related Exhaust... fuel is used for all the duty cycles specified in 40 CFR part 86, subpart S. If a vehicle fails to meet... compliance with all applicable emission standards over that test cycle. (c) If a manufacturer...

  13. Natural tooth as an interim prosthesis

    PubMed Central

    Dhariwal, Neha S.; Gokhale, Niraj S.; Patel, Punit; Hugar, Shivayogi M.

    2016-01-01

    A traumatic injury to primary maxillary anterior tooth is one of the common causes for problems with the succedaneous tooth leading to it noneruption. A missing anterior tooth can be psychologically and socially damaging to the patient. Despite a wide range of treatment options available, sometimes, it is inevitable to save the natural tooth. This paper describes the immediate replacement of a right central incisor using a fiber-composite resin splint with the natural tooth crown as a pontic following surgical extraction of the dilacerated impacted permanent maxillary central incisor. The abutment teeth can be conserved with minimal or no preparation, thus keeping the technique reversible and can be completed at chair side thereby avoiding laboratory costs. It can be used as an interim measure until a definitive prosthesis can be fabricated as the growth is still incomplete. PMID:27433074

  14. Interim prediction method for jet noise

    NASA Technical Reports Server (NTRS)

    Stone, J. R.

    1974-01-01

    A method is provided for predicting jet noise for a wide range of nozzle geometries and operating conditions of interest for aircraft engines. Jet noise theory, data and existing prediction methods was reviewed, and based on this information a interim method of jet noise prediction is proposed. Problem areas are idenified where further research is needed to improve the prediction method. This method predicts only the noise generated by the exhaust jets mixing with the surrounding air and does not include other noises emanating from the engine exhaust, such as combustion and machinery noise generated inside the engine (i.e., core noise). It does, however, include thrust reverser noise. Prediction relations are provided for conical nozzles, plug nozzles, coaxial nozzles and slot nozzles.

  15. High Temperature Materials Interim Data Qualification Report FY 2011

    SciTech Connect

    Nancy Lybeck

    2011-08-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim fiscal year (FY) 2011 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under the Nuclear Quality Assurance (NQA)-1 guidelines and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from seven test series within the High Temperature Materials data stream have been entered into the NDMAS vault, including tensile tests, creep tests, and cyclic tests. Of the 5,603,682 records currently in the vault, 4,480,444 have been capture passed, and capture testing is in process for the remaining 1,123,238.

  16. CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

    SciTech Connect

    Vinson, D.

    2010-07-11

    Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a 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. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and at cask

  17. Updated NGNP Fuel Acquisition Strategy

    SciTech Connect

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01

    A Next Generation Nuclear Plant (NGNP) fuel acquisition strategy was first established in 2007. In that report, a detailed technical assessment of potential fuel vendors for the first core of NGNP was conducted by an independent group of international experts based on input from the three major reactor vendor teams. Part of the assessment included an evaluation of the credibility of each option, along with a cost and schedule to implement each strategy compared with the schedule and throughput needs of the NGNP project. While credible options were identified based on the conditions in place at the time, many changes in the assumptions underlying the strategy and in externalities that have occurred in the interim requiring that the options be re-evaluated. This document presents an update to that strategy based on current capabilities for fuel fabrication as well as fuel performance and qualification testing worldwide. In light of the recent Pebble Bed Modular Reactor (PBMR) project closure, the Advanced Gas Reactor (AGR) fuel development and qualification program needs to support both pebble and prismatic options under the NGNP project. A number of assumptions were established that formed a context for the evaluation. Of these, the most important are: • Based on logistics associated with the on-going engineering design activities, vendor teams would start preliminary design in October 2012 and complete in May 2014. A decision on reactor type will be made following preliminary design, with the decision process assumed to be completed in January 2015. Thus, no fuel decision (pebble or prismatic) will be made in the near term. • Activities necessary for both pebble and prismatic fuel qualification will be conducted in parallel until a fuel form selection is made. As such, process development, fuel fabrication, irradiation, and testing for pebble and prismatic options should not negatively influence each other during the period prior to a decision on reactor type

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

    SciTech Connect

    JOHNSON, D.M.

    2000-01-06

    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.

  19. Interim Guidance: Municipal Solid Waste Exemption under Superfund

    EPA Pesticide Factsheets

    Interim guidance discusses the statutory provisions of CERCLA § 107(p) and identifies some factors to be considered by EPA and DOJ staff in exercising their enforcement discretion of MSW at an NPL site.

  20. 46 CFR 308.303 - Amounts insured under interim binder.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... INSURANCE Second Seamen's War Risk Insurance § 308.303 Amounts insured under interim binder. The amounts insured are the amounts specified in the Second Seamen's War Risk Policy (1955) or as modified by...

  1. 46 CFR 308.303 - Amounts insured under interim binder.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... INSURANCE Second Seamen's War Risk Insurance § 308.303 Amounts insured under interim binder. The amounts insured are the amounts specified in the Second Seamen's War Risk Policy (1955) or as modified by...

  2. 46 CFR 308.303 - Amounts insured under interim binder.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... INSURANCE Second Seamen's War Risk Insurance § 308.303 Amounts insured under interim binder. The amounts insured are the amounts specified in the Second Seamen's War Risk Policy (1955) or as modified by...

  3. 46 CFR 308.303 - Amounts insured under interim binder.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... INSURANCE Second Seamen's War Risk Insurance § 308.303 Amounts insured under interim binder. The amounts insured are the amounts specified in the Second Seamen's War Risk Policy (1955) or as modified by...

  4. 40 CFR 155.56 - Interim registration review decision.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... required data, conducting the new risk assessment and completing the registration review. A FIFRA 3(c)(2)(B... registration review decision may require new risk mitigation measures, impose interim risk mitigation...

  5. 40 CFR 155.56 - Interim registration review decision.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... required data, conducting the new risk assessment and completing the registration review. A FIFRA 3(c)(2)(B... registration review decision may require new risk mitigation measures, impose interim risk mitigation...

  6. 40 CFR 155.56 - Interim registration review decision.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... required data, conducting the new risk assessment and completing the registration review. A FIFRA 3(c)(2)(B... registration review decision may require new risk mitigation measures, impose interim risk mitigation...

  7. 40 CFR 155.56 - Interim registration review decision.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... required data, conducting the new risk assessment and completing the registration review. A FIFRA 3(c)(2)(B... registration review decision may require new risk mitigation measures, impose interim risk mitigation...

  8. 14 CFR 91.871 - Waivers from interim compliance requirements.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... period will be at least 14 days. ... 14 Aeronautics and Space 2 2010-01-01 2010-01-01 false Waivers from interim compliance... TRANSPORTATION (CONTINUED) AIR TRAFFIC AND GENERAL OPERATING RULES GENERAL OPERATING AND FLIGHT RULES...

  9. 45 CFR 689.8 - Interim administrative actions.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... RESEARCH MISCONDUCT § 689.8 Interim administrative actions. (a) After an inquiry or during an external or... misconduct. Such an order will normally be issued on recommendation from OIG and in consultation with...

  10. TANK FARM INTERIM SURFACE BARRIER MATERIALS AND RUNOFF ALTERNATIVES STUDY

    SciTech Connect

    HOLM MJ

    2009-06-25

    This report identifies candidate materials and concepts for interim surface barriers in the single-shell tank farms. An analysis of these materials for application to the TY tank farm is also provided.

  11. EPA's Revised Interim Financial Assistance Conflict of Interest Policy

    EPA Pesticide Factsheets

    EPA has established the following revised interim policy governing disclosure of actual and potential conflicts of interest (COI Policy) by applicants for, and recipients of, federal financial assistance awards from EPA.

  12. The growing problem of stranded used nuclear fuel.

    PubMed

    Alley, William M; Alley, Rosemarie

    2014-02-18

    By 2050, almost all U.S. nuclear reactors will have reached their 60 year maximum expected life. Many will shut down sooner. With no assurance that the current approach for finding a geologic repository or interim storage sites will succeed, used nuclear fuel could be stranded indefinitely at more than 70 sites in 35 states. Societal discussions about the future of nuclear waste should be framed in terms of the relative risks of all alternatives. We review and compare onsite storage, interim storage, and a geologic repository, as well as how these alternatives are presented to the public.

  13. Fossil fuels -- future fuels

    SciTech Connect

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  14. Processing of driver fuel assemblies at FFTF

    SciTech Connect

    Danko, A.D.; Hicks, D.F.; Arneson, S.O.

    1982-07-01

    The ability to disassemble an irradiated Fast Flux Test Facility (FFTF) Driver Fuel Assembly (DFA) is important both to the continued operation of the FFTF and the future of the Breeder Reactor Program. At the FFTF, DFA's with up to three (3)* kilowatts of decay heat will be placed in the Interim Examination and Maintenance (IEM) Cell for disassembly and nondestructive examination. This process includes sodium removal, duct measurement, duct cutting and pulling, fuel pin removal, and component disposition to other laboratories for destructive examination.

  15. K basins interim remedial action health and safety plan

    SciTech Connect

    DAY, P.T.

    1999-09-14

    The K Basins Interim Remedial Action Health and Safety Plan addresses the requirements of the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), as they apply to the CERCLA work that will take place at the K East and K West Basins. The provisions of this plan become effective on the date the US Environmental Protection Agency issues the Record of Decision for the K Basins Interim Remedial Action, currently planned in late August 1999.

  16. An approach to determine a defensible spent fuel ratio.

    SciTech Connect

    Durbin, Samuel G.; Lindgren, Eric Richard

    2014-03-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. However, the validity of these large-scale results remain in question due to the lack of a defensible Spent Fuel Ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980s and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000s. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort

  17. Conservation trend indicators. Interim and final report

    SciTech Connect

    Peterson, C.

    1980-08-21

    This report outlines major conservation trends for the four US energy consuming sectors: residential, commercial, industrial, and transportation. Tables on residential use include data on gas, electricity, and fuel oil. Data is not available for renewable sources or for bottled gas, kerosene, or propane. Commercial data is also reported for gas, electricity, and fuel oil, and total consumption is examined on a square foot basis. Industrial trends are shown for the ten most energy intensive industries. In addition, industrial efficiencies by fuel are analyzed. For the transportation sector, petroleum products (particularly gasoline) are the major fuels considered, and selected energy intensities are examined. The discussion on each sector will cover findings, data sources, and data interpretation. Recommendations on reporting are part of the residential section of this report.

  18. 40 CFR 92.12 - Interim provisions.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... specified in § 92.212(c)(2)(v). (i) Diesel test fuels. Manufacturers and remanufacturers may use LSD or ULSD... determine your cycle-weighted emission rates as specified in subpart B of this part. If you test using...

  19. 40 CFR 92.12 - Interim provisions.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... specified in § 92.212(c)(2)(v). (i) Diesel test fuels. Manufacturers and remanufacturers may use LSD or ULSD... determine your cycle-weighted emission rates as specified in subpart B of this part. If you test using...

  20. 40 CFR 92.12 - Interim provisions.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... specified in § 92.212(c)(2)(v). (i) Diesel test fuels. Manufacturers and remanufacturers may use LSD or ULSD... determine your cycle-weighted emission rates as specified in subpart B of this part. If you test using...