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Sample records for fuel dry interim

  1. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect

    Romanato, L.S.

    2008-07-01

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

  2. 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.

  3. 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.

  4. 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.

  5. Study on Hydride Reorientation in Zry-2 Fuel Claddings during Interim Dry Storage

    SciTech Connect

    Sakamoto, K.; Matsuoka, H.; Takagi, A.; Kashibe, S.

    2007-07-01

    The hydride reorientation during the interim dry storage was examined by hydride reorientation test using unirradiated recrystallized Zry-2 fuel claddings (Zr-lined). In the case of high hydrogen concentration (above 200 ppm), no measurable hydride reorientation was observed under the condition examined. On the other hand, for low hydrogen concentration (30 - 80 ppm), a significant hydride reorientation was observed above 618 K. The effects of thermal cycling and cooling rate were also examined. The mechanical property of the hydride-reoriented specimens was evaluated at room temperature by the ring-tensile test, which showed no degradation of hoop strength and ductility when temperature and hoop stress were not greater than 573 K and 70 MPa, even if the effects of cooling rate and thermal cycling were taken into account. (authors)

  6. Hanford`s progress toward dry interim storage of K basin`s spent fuel

    SciTech Connect

    Culley, G.E., Westinghouse Hanford

    1996-05-09

    This paper highlights the progress made toward removing the U.S. Department of Energy`s (DOE) approximately 2, 100 metric tons of metallic spent nuclear fuel from two outdated K Basins on the banks of the Columbia River and placing it in safe, economic interim dry storage beginning in December 1997. A new way of doing business at the Hanford Site and within DOE is being used to achieve the fast-track schedule, , cost savings, and public cooperation needed for success. In February 1994, the Spent Nuclear Fuel (SNF) Project was formed to solve serious safety and environmental problems associated with corroding metallic spent fuel stored in 1950`s vintage, leak-prone, water- filled concrete basins located within 365 meters (400 yards) of the last remaining unspoiled section of the Columbia River. Working together, the integrated project team focused on quickly getting the fuel out of the basins and into safe, dry storage. The team involved the public, government, regulators, and other stakeholders and forged a common understanding. The DOE transferred authority to the field to shorten approval times, and Site contractors reengineered processes to improve efficiency. Within nine months of creating the project, a plan was recommended to the DOE. It was approved on February 14, 1995. Further refinement, during the following six months, shortened the schedule even more and reduced costs by $350 million. The SNF Project is on a fast track. The K Basins Environmental Impact Statement was completed in only 11 months for only $1.3 million. Fuel and sludge samples were obtained from both basins and were sent to the laboratory for characterization and testing. The partially constructed Canister Storage Building (CSB), selected as the fuel storage facility, was redesigned, and construction was restarted saving over $17 million and cutting a year off the project schedule. With fuel removal beginning in December 1997, the SNF Project will have the fuel out of the K Basins and into

  7. 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.

  8. Dosimetry at an interim storage for spent nuclear fuel.

    PubMed

    Králík, M; Kulich, V; Studeny, J; Pokorny, P

    2007-01-01

    The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons. PMID:17526479

  9. 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.

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

    SciTech Connect

    Grenier, R.M.; Koploy, M.A.

    1995-12-31

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

  11. Equipment designs for the spent LWR fuel dry storage demonstration

    SciTech Connect

    Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

    1980-01-01

    In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations.

  12. Fuel Supply Shutdown Facility Interim Operational Safety Requirements

    SciTech Connect

    BENECKE, M.W.

    2000-09-06

    The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility.

  13. 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.

  14. Report on interim storage of spent nuclear fuel

    SciTech Connect

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  15. 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

  16. Fuel supply shutdown facility interim operational safety requirements

    SciTech Connect

    Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

    1995-05-23

    These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance).

  17. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    SciTech Connect

    Richard, R.F.

    1995-05-11

    It has been postulated that a degradation phenomenon, referred to as ``hot cell rot``, may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ``Hot cell rot`` refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ``hot cell rot`` phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical.

  18. 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.

  19. International status of dry storage of spent fuels

    SciTech Connect

    Schneider, K.J.; Mitchell, S.J.; Johnson, A.B. Jr.

    1992-04-01

    Spent fuel from the world`s nuclear power reactors, or the high-level radioactive wastes from reprocessing of the spent fuels, are planned to be disposed of in national deep geological repositories in the respective countries of origin. The plans for most countries with nuclear power call for spent fuel or high-level waste disposal to start between 2010 and about 2050. Although storage in water pools is the primary method for management of spent nuclear fuels for the first few years after discharge from the reactor, dry storage has been implemented in several countries and is being considered in others. Dry storage is generally planned for an interim period (from 10 to as long as 100 years) until the spent fuel is disposed of or until a final decision is made on reprocessing. Dry storage is also being used to supplement wet storage capacity at some nuclear power stations. This paper summarizes the world-wide status of dry spent fuel storage and information on the expected long-term integrity of the dry-stored spent fuel based on experience, particularly for Zircaloy-clad fuels. The paper also addresses briefly the dry storage of solidified high-level radioactive wastes. This paper is based on work carried out for the US Department of Energy (DOE) by the Pacific Northwest Laboratory.

  20. International status of dry storage of spent fuels

    SciTech Connect

    Schneider, K.J.; Mitchell, S.J.; Johnson, A.B. Jr.

    1992-04-01

    Spent fuel from the world's nuclear power reactors, or the high-level radioactive wastes from reprocessing of the spent fuels, are planned to be disposed of in national deep geological repositories in the respective countries of origin. The plans for most countries with nuclear power call for spent fuel or high-level waste disposal to start between 2010 and about 2050. Although storage in water pools is the primary method for management of spent nuclear fuels for the first few years after discharge from the reactor, dry storage has been implemented in several countries and is being considered in others. Dry storage is generally planned for an interim period (from 10 to as long as 100 years) until the spent fuel is disposed of or until a final decision is made on reprocessing. Dry storage is also being used to supplement wet storage capacity at some nuclear power stations. This paper summarizes the world-wide status of dry spent fuel storage and information on the expected long-term integrity of the dry-stored spent fuel based on experience, particularly for Zircaloy-clad fuels. The paper also addresses briefly the dry storage of solidified high-level radioactive wastes. This paper is based on work carried out for the US Department of Energy (DOE) by the Pacific Northwest Laboratory.

  1. Studies and research concerning BNFP: spent fuel dry storage studies at the Barnwell Nuclear Fuel Plant

    SciTech Connect

    Anderson, Kenneth J.

    1980-09-01

    Conceptual designs are presented utilizing the Barnwell Nuclear Fuel Plant for the dry interim storage of spent light water reactor fuel. Studies were conducted to determine feasible approaches to storing spent fuel by methods other than wet pool storage. Fuel that has had an opportunity to cool for several years, or more, after discharge from a reactor is especially adaptable to dry storage since its thermal load is greatly reduced compared to the thermal load immediately following discharge. A thermal analysis was performed to help in determining the feasibility of various spent fuel dry storage concepts. Methods to reject the heat from dry storage are briefly discussed, which include both active and passive cooling systems. The storage modes reviewed include above and below ground caisson-type storage facilities and numerous variations of vault, or hot cell-type, storage facilities.

  2. 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

  3. Foreign experience in extended dry storage of spent nuclear fuel

    SciTech Connect

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

    1992-06-01

    Most countries with nuclear power are planning for spent nuclear fuel (or high-level waste from reprocessing of spent fuel) to be disposed of in national deep geological repositories starting in the time period of about 2010 to 2050. While spent fuel has been stored in water basins for the early years after discharge from the reactors, interim dry storage for extended periods (i.e., several tens of years) is being implemented or considered in an increasing number of countries. Dry storage technology is generally considered to be developed on a world-wide basis, and is being initiated and/ or expanded in a number of countries. This paper presents a summary of status and experience in dry storage of spent fuel in other countries, with emphasis on zirconium-clad fuels. Past activities, current status, future plans, research and development, and experience in dry storage are summarized for Argentina, Canada, France, former West Germany, former East Germany, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former Soviet Union. Conclusions from their experience are presented. Their experience to date supports the expectations that proper dry storage should provide for safe extended dry storage of spent fuel.

  4. 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.

  5. Dry process dependency of dupic fuel cycle

    SciTech Connect

    Park, Kwangheon; Whang, Juho; Kim, Yun-goo; Kim, Heemoon

    1996-12-31

    During the Dry Process, volatile and semi-volatile elements are released from the fuel. The effects of these released radioactive nuclides on DUPIC fuel cycle are analyzed from the view-point of radiation hazard, decay beat, and hazard index. Radiation hazard of fresh and spent DUPIC fuel is sensitive to the method of Dry Process. Decay beat of the fuel is also affected. Hazard index turned out not to be dependent on Dry Process.

  6. 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.

  7. 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...

  8. 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.

  9. 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.

  10. 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).

  11. 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.

  12. 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.

  13. 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.

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

    SciTech Connect

    Masumi Wataru; Hisashi Kato; Satoshi Kudo; Naoko Oshima; Koji Wada; Hirofumi Narutaki

    2006-07-01

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

  15. 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.

  16. Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

    SciTech Connect

    Dawson, D.M.; Guerra, G.; Neider, T.; Shih, P.

    1995-12-01

    This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity or other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.

  17. 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.

  18. 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.

  19. 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).

  20. Hog fuel drying using vapour recompression

    SciTech Connect

    Azarniouch, M.K.; Sayegh, N.N.

    1983-12-01

    Hog fuel is a broad term used by the forest products industry to describe all types of wood residues that are used as fuel. These can be bark, wood chip rejects, sawdust, shavings, etc. The moisture content of these residuals depends mainly on their source and on the conditions of storage and could range from 50 to 70% (wet basis). The introduction of the significant amounts of water contained in the hog fuel impairs the efficiency of operating a hog-fuel boiler by: (i) reducing the net heating value of the fuel, (ii) increasing the particulate emissions caused by incomplete combustion, (iii) requiring higher air flows to maintain combustion, (iv) decreasing the rate of combustion, and (v) reducing the heat flux as a result of lower flame temperatures. Several drying systems are commercially available for the drying of hog fuel and most of them use the waste heat in the flue gas as the source of heat for the drying operation. Commercial flue gas dryers are rotary, cascade or flash type. However, all of these suffer from the same inherent disadvantages.

  1. 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.

  2. Assessment of nitrogen as an atmosphere for dry storage of spent LWR fuel

    SciTech Connect

    Gilbert, E.R.; Knox, C.A.; White, G.D.

    1985-09-01

    Interim dry storage of spent light-water reactor (LWR) fuel is being developed as a licensed technology in the United States. Because it is anticipated that license agreements will specify dry storage atmospheres, the behavior of spent LWR fuel in a nitrogen atmosphere during dry storage was investigated. In particular, the thermodynamics of reaction of nitrogen compounds (expected to form in the cover gas during dry storage) and residual impurities (such as moisture and oxygen) with Zircaloy cladding and with spent fuel at sites of cladding breaches were examined. The kinetics of reaction were not considered it was assumed that the 20 to 40 years of interim dry storage would be sufficient for reactions to proceed to completion. The primary thermodynamics reactants were found to be NO/sub 2/, N/sub 2/O, H/sub 2/O/sub 2/, and O/sub 2/. The evaluation revealed that the limited inventories of these reactants produced by the source terms in hermetically sealed dry storage systems would be too low to cause significant spent fuel degradation. Furthermore, the oxidation of spent fuel to degrading O/U ratios is unlikely because the oxidation potential in moist nitrogen limits O/U ratios to values less than UO/sub 2.006/ (the equilibrium stoichiometric form in equilibrium with moist nitrogen). Tests were performed with bare spent UO/sub 2/ fuel and nonirradiated UO/sub 2/ pellets (with no Zircaloy cladding) in a nitrogen atmosphere containing moisture concentrations greater than encountered under dry storage conditions. These tests were performed for at least 1100 h at temperatures as high as 380/sup 0/C, where oxidation reactions proceed in a matter of minutes. No visible degradation was detected, and weight changes were negligible.

  3. 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.

  4. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    SciTech Connect

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  5. 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

    ... On July 10, 2009, notice was given in the Federal Register (74 FR 33281) of the availability for... COMMISSION Notice of Availability of Interim Staff Guidance Document for Fuel Cycle Facilities AGENCY.... Powell, Nuclear Process Engineer, Technical Support Branch, Division of Fuel Cycle Safety and...

  6. Interim safety basis for fuel supply shutdown facility

    SciTech Connect

    Brehm, J.R.; Deobald, T.L.; Benecke, M.W.; Remaize, J.A.

    1995-05-23

    This ISB in conjunction with the new TSRs, will provide 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. It is concluded that the risk associated with the current operational mode of the Facility, uranium closure, clean up, and transition activities required for permanent closure, are within Risk Acceptance Guidelines. The Facility is classified as a Moderate Hazard Facility because of the potential for an unmitigated fire associated with the uranium storage buildings.

  7. 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.

  8. Temperature for Spent Fuel Dry Storage

    SciTech Connect

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimates of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.

  9. Temperature for Spent Fuel Dry Storage

    Energy Science and Technology Software Center (ESTSC)

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimatesmore » of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.« less

  10. 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.

  11. 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.

  12. 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.

  13. Drying studies for corroded DOE aluminum plate fuels

    SciTech Connect

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

    1996-05-01

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

  14. 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.

  15. 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.

  16. 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.

  17. Criteria for Corrosion Protection of Aluminum-Clad Spent Nuclear Fuel in Interim Wet Storage

    SciTech Connect

    Howell, J.P.

    1999-09-14

    Storage of aluminum-clad spent nuclear fuel at the Savannah River Site (SRS) and other locations in the U. S. and around the world has been a concern over the past decade because of the long time interim storage requirements in water. Pitting corrosion of production aluminum-clad fuel in the early 1990''s at SRS was attributed to less than optimum quality water and corrective action taken has resulted in no new pitting since 1994. The knowledge gained from the corrosion surveillance testing and other investigations at SRS over the past 8 years has provided an insight into factors affecting the corrosion of aluminum in relatively high purity water. This paper reviews some of the early corrosion issues related to aluminum-clad spent fuel at SRS, including fundamentals for corrosion of aluminum alloys. It updates and summarizes the corrosion surveillance activities supporting the future storage of over 15,000 research reactor fuel assemblies from countries over the world during the next 15-20 years. Criteria are presented for providing corrosion protection for aluminum-clad spent fuel in interim storage during the next few decades while plans are developed for a more permanent disposition.

  18. 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.

  19. The Hanford spent nuclear metal fuel multi-canister overpack and vacuum drying {ampersand} hot conditioning process

    SciTech Connect

    Irwin, J.J.

    1996-05-15

    Nuclear production reactors operated at the U.S. Department of Energy`s Hanford Site from 1944 until 1988 to produce plutonium. Most of the irradiated fuel from these reactors was processed onsite to separate and recover the plutonium. When the processing facilities were closed in 1992, about 1,900 metric tons of unprocessed irradiated fuel remained in storage. Additional fuel was irradiated for research purposes or was shipped to the Hanford Site from offsite reactor facilities for storage or recovery of nuclear materials. The fuel inventory now in storage at the Hanford Site is predominantly N Reactor irradiated fuel, a metallic uranium alloy that is coextruded into zircaloy-2 cladding. The Spent Nuclear Fuel Project has rommitted to an accelerated schedule for removing spent nuclear fuel from the Hanford Site K Basins to a new interim storage facility in the 200 Area. Under the current proposed accelerated schedule, retrieval of spent nuclear fuel stored in the K East and West Basins must begin by December 1997 and be completed by December 1999. A key part of this action is retrieving fuel canisters from the water-filled K Basin storage pools and transferring them into multi@ister overpacks (MCOS) that will be used to handle and process the fuel, then store it after conditioning. The Westinghouse Hanford Company has developed an integrated process to deal with the K Basin spent fuel inventory. The process consists of cleaning the fuel, packaging it into MCOS, vacuum drying it at the K Basins, then transporting it to the Canister Storage Building (CSB) for staging, hot conditioning, and interim storage. This presentation dekribes the MCO function, design, and life-cycle, including an overview of the vacuum drying and hot conditioning processes.

  20. 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.

  1. 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.

  2. 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

  3. Storing the Spent Nuclear Fuel in Dry Casks Licensed for a Century as an Alternative to Recycling Solution

    NASA Astrophysics Data System (ADS)

    Milincic, Radovan

    2010-02-01

    Management of spent nuclear power reactor fuels is one of the most urgent problems in nuclear technology. Yearly production of new spent fuel is in the range of thousands of tons, topping a couple of hundred thousand tons of spent fuel already. This material is extremely radioactive and currently there is no adequate international policy, control or management regarding it. I propose here an intermediate term solution to this problem, which will be technologically and economically sustainable: interim spent-fuel storage as an alternative to reprocessing. The reprocessing inherently increases the net amount of the plutonium, which can be used for production of nuclear arms. Moreover, it is an expensive process with the net effect of producing different type of radioactive waste. In particular, the development of a dry cask for nuclear waste storage on site and transport, licensed for a period of hundred years would provide a significantly less expensive solution in the recent future, giving a needed relief to crowded spent-fuel storage pools. Currently in the U.S, NRC licenses existing storage casks for 20 years; and licenses for some of the dry cask storage facilities in the U.S. are about to expire. The extended life dry casks will provide sufficient intermediate period toward a more efficient and/or technologically advanced solution for spent fuel. )

  4. 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

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

    SciTech Connect

    Einziger, R. E.

    1998-12-16

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and

  6. Spent fuel storage: Progress with modular vault dry storage

    SciTech Connect

    Bower, C.C.F.

    1995-12-31

    This paper discusses the Modular Vault Dry Store (MVDS) for spent fuels at the Wylfa nuclear power plant in North Wales and at Fort St Vrain in Colorado. It goes on to discuss Scottish Nuclear`s decision not to proceed with MVDS facilities. It concludes by discussing Paks NPP contract with GEC Alsthom for the design and safety case for MDVS.

  7. 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.

  8. 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.

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

    NASA Astrophysics Data System (ADS)

    Casella, Andrew M.

    2007-09-01

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

  10. An information management system for a spent nuclear fuel interim storage facility.

    SciTech Connect

    Finch, Robert J.; Chiu, Hsien-Lang; Giles, Todd; Horak, Karl Emanuel; Jow, Hong-Nian

    2010-12-01

    We describe an integrated information management system for an independent spent fuel dry-storage installation (ISFSI) that can provide for (1) secure and authenticated data collection, (2) data analysis, (3) dissemination of information to appropriate stakeholders via a secure network, and (4) increased public confidence and support of the facility licensing and operation through increased transparency. This information management system is part of a collaborative project between Sandia National Laboratories, Taiwan Power Co., and the Fuel Cycle Materials Administration of Taiwan's Atomic Energy Council, which is investigating how to implement this concept.

  11. 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)

  12. 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.

  13. 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.

  14. 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).

  15. 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.

  16. 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.

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

    SciTech Connect

    Unterzuber, R.; Wright, J.B.

    1980-09-01

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

  18. 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

  19. 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.

  20. 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.

  1. 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)

  2. Reaction rate constant for dry air oxidation of K Basin fuel

    SciTech Connect

    Trimble, D.J.

    1998-04-29

    The rate of oxidation of spent nuclear fuel stored in the K Basin water is an important parameter when assessing the processes and accident scenarios for preparing the fuel for dry storage. The literature provides data and rate laws for the oxidation of unirradiated uranium in various environments. Measurement data for the dry air oxidation of K Basin fuel is compared to the literature data for linear oxidation in dry air. Equations for the correlations and statistical bounds to the K Basin fuel data and the literature data are selected for predicting nominal and bounding rates for the dry air oxidation of the K Basin fuel. These rate equations are intended for use in the Spent Nuclear Fuel Project Technical Data book.

  3. 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.

  4. 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.

  5. 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

  6. Biodiesel fuel technology for military application. Interim report, July 1994-May 1996

    SciTech Connect

    Frame, E.A.; Bessee, G.B.; Marbach, H.W.

    1997-12-01

    This program addressed the effects of biodiesel (methyl soyate) and blends of biodiesel with petrofuels on fuel system component and material compatibility, fuel storage stability, and fuel lubricity. Biodiesel was found to have excellent lubricity properties and was effective at 1 volume percent (vol %) blend in improving the lubricity of Jet A-1 fuel. The following potential problem areas associated with methyl soyate use were identified: storage stability, compatibility with some metals, and compatibility with nitrile elastomers.

  7. Navy fuel-specification standardization. Interim report, Jul 86-May 87

    SciTech Connect

    Tosh, J.D.; Moulton, D.S.; Moses, C.A.

    1992-04-01

    In the early 1970's, the U.S. Navy switched from use of Navy Special Fuel Oil (NSFO NATO F-77 to Naval Distillate Fuel NATO F-76) in shipboard propulsion and electric generating systems. Presently there are two fuels being utilized in shipboard operations. NDF (F-76) is utilized in propulsion and electric generating systems, and JP-5 jet fuel is used for aircraft operations and as an emergency fuel for systems utilizing F-76. Since the conversion from NSFO to F-76, there has been interest in an additional conversion to a shipboard single-fuel operation. Due to the unique requirement of jet aircraft engines, single fuel, from necessity, would have to be JP-5. Therefore, this study was conducted to determine the potential benefits and problems associated with a shipboard single-fuel operation. All shipboard systems, including boilers, turbine engines, and diesel engines should continue to operate satisfactorily, and in some instances, with increased efficiency with JP-5. The greatest benefit from such a conversion would be the convenience of handling only one fuel, and eliminating the possibility of fuel contamination. The major penalties would include higher fuel cost, and difficulty in procuring adequate supplies of JP-5 to meet the total U.S. Navy shipboard fuel requirements.

  8. Fuel-lubricity requirements for diesel-injection systems. Interim report, Sep 90-Feb 91

    SciTech Connect

    Lacey, P.I.; Lestz, S.J.

    1991-02-01

    The U.S. Department of Defense has adopted the single fuel for the battlefield concept. Diesel fuel will be replace by JP-8/Jet A-1, which has both lower lubricity and viscosity. Currently, the tribological requirements of fuel-lubricated components in the injection system are unknown. As a result, no widely approved lubricity test or standard exists. Similar problems are currently faced in commercial applications where low-sulfur/aromatic fuels are being introduced. The present study details the wear mechanisms likely to exist with low lubricity fuels, with particular reference to injection equipment known to be fuel sensitive. The wear mechanism was found to ba a function of contact severity and may not be uniquely defined by a single test. A number of potentially viable lubricity tests is suggested, and fuel/additive components are recommended for wear reduction.

  9. Thermal-oxidation stability of diesel fuels. Interim report, October 1983-January 1986

    SciTech Connect

    Stavinoha, L.L.; Barbee, J.G.; Yost, D.M.

    1986-02-01

    Injector fouling bench tests(IFBT) and modified Jet Fuel Thermal Oxidation Test(JFTOT, ASTMD 3241) were used to develop methodology for evaluating the thermal stability of diesel fuels. A new method for measuring the thickness of lacquer-type fuel deposits formed on test surfaces at elevated temperatures was developed and applied to a variety of fuels, both with and without MIL-S-53021(additive stabilizer package). The utility of this technique greatly expands the capability for exploring and defining diesel-fuel thermal stability with respect to both material and kinetic studies. Correlation of IFBT and JFTOT type tests including definitions of temperature, flow, test-surface metallurgy and fuel additive effects can now be performed to better understand diesel thermal stability and provide test methodology/test limit information for fuel-specification consideration.

  10. Fuel requirements for low-heat rejection military diesel engines. Interim report, October 1991-September 1993

    SciTech Connect

    Westbrook, S.R.; Stavioha, L.L.; McInnis, L.A.; Likos, W.E.; Yost, D.M.

    1996-01-01

    In the development of high-efficiency advanced engine technology such as low-heat rejection engines and injection systems, the thermal stability of fuel is an important concern. The next generation of engines for combat vehicles will be operating at higher fuel temperatures due to lower waste heat rejection and will be accompanied by higher heat transfer to the fuel injection system. Thus, high-temperature fuel deposit formation is more likely. As a result, two possible methods were evaluated for their potential to reduce fuel deposits: (1) prestress the fuel in an apparatus that feeds the fuel to the engine, or (2) pretreat the fuel with an appropriate additive to reduce deposits in the engine. It was shown that removal of dissolved oxygen from the fuel can significantly reduce the formation of deposits on hot metal surfaces. Prestressing the fuel prior to burning it in the engine was also effective in the reduction of deposit formation. The use of additive pretreatment yielded only limited success.

  11. 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%.

  12. 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...

  13. 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.

  14. Resource recovery of organic sludge as refuse derived fuel by fry-drying process.

    PubMed

    Chang, Fang-Chih; Ko, Chun-Han; Wu, Jun-Yi; Wang, H Paul; Chen, Wei-Sheng

    2013-08-01

    The organic sludge and waste oil were collected from the industries of thin film transistor liquid crystal display and the recycled cooking oil. The mixing ratio of waste cooking oil and organic sludge, fry-drying temperatures, fry-drying time, and the characteristics of the organic sludge pellet grain were investigated. After the fry-drying process, the moisture content of the organic sludge pellet grain was lower than 5% within 25 min and waste cooking oil was absorbed on the dry solid. The fry-drying organic sludge pellet grain was easy to handle and odor free. Additionally, it had a higher calorific value than the derived fuel standards and could be processed into organic sludge derived fuels. Thus, the granulation and fry-drying processes of organic sludge with waste cooking oil not only improves the calorific value of organic sludge and becomes more valuable for energy recovery, but also achieves waste material disposal and cost reduction. PMID:23623433

  15. WET/DRY COOLING SYSTEMS FOR FOSSIL-FUELED POWER PLANTS: WATER CONSERVATION AND PLUME ABATEMENT

    EPA Science Inventory

    The report gives results of a study of technical and economic feasibilities of wet/dry cooling towers for water conservation and vapor plume abatement. Results of cost optimizations of wet/dry cooling for 1000-MWe fossil-fueled power plants are presented. Five sites in the wester...

  16. 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.…

  17. Improved drying rate diagnostics for saturated fuel debris at the INEEL

    SciTech Connect

    Childs, K.; Christensen, A.

    1999-09-01

    A fuel canning station (FCS) has been operated for {approximately}2 yr to prepare for the dry storage of a variety of spent reactor fuels stored in pools at the Idaho National Engineering and Environmental Laboratory (INEEL). The FCS dewaters the fuel and then passivates possibly pyrophoric components in the fuel. Fuel-loaded canisters are placed into a heated insert, the canister is connected to a vacuum system, and the fuel is heated under a vacuum to remove the water. The dewatering system must also verify that the water was removed. The dryness criteria state that the canister pressure shall not exceed a defined pressure for a specified isolation time. Dewatering did not work well for defected TRIGA elements that had corroded in pool storage, leaving the intact fuel meat mixed with a bed of fines from metal oxides and from sludge that continuously accumulated within the pool. Dewatering these cans proved to be very time consuming. Fueled canisters were heated to 60 C and evacuated between 5 and 10 torr. At these conditions, intact fuels were rapidly dried (<10 h). TRIGA drying periods extended to 9 days. Dryness was qualitatively monitored using the canister pressure-control valve position. The valve closes as the gas flow rate declines, providing an indication that drying is complete. However, the valve remained open when drying TRIGA fuel, leaving no indication of dryness. In addition, dryness could not be verified because the canister pressure exceeded the defined pressure during isolation. Air leakage into the evacuated canister prevented the dryness from being verified. Air in-leakage and water vapor cannot easily be discriminated by the aforementioned procedures. Because the canister design does not seal above atmospheric pressure, a drying temperature that yielded a vapor pressure less than atmospheric pressure was chosen. A sufficiently long isolation test could then determine if air was accumulating in the canister; however, the low temperature reduced

  18. 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)

  19. 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 ...

  20. 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

  1. Spent fuel test-climax: technical measurements interim report, FY 1980

    SciTech Connect

    Carlson, R.C.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Montan, D.N.; Harben, P.E.; Ballou, L.B.; Heard, H.C.

    1980-12-01

    The Spent Fuel Test--Climax (SFT-C), a test of the retrievable geologic storage of spent fuel assemblies from an operating commercial power reactor, is under way at the Nevada Test Site of the US Department of Energy. Although the main thrust of the project is a demonstration of the feasibility of packaging, handling, storing, and retrieving the highly radioactive fuel assemblies, over 800 data channels have been installed to monitor the response of the rock to the heat and radiation produced by the fuel assemblies and to distinguish in that response the effect due to heat alone. Temperatures in the test array are tracking well with thermal modeling calculations performed before the test was started. The fuel assemblies have been in place since May 1980. The canisters have passed through skin temperature maxima of about 145/sup 0/C and are currently declining in temperature. Evidence is emerging that the thermomechanical response of the rock surrounding the SFT-C is strongly affected by fractures and other discontinuities inthe rock. Most of the effort to date has been in project construction, design, and installation of the instrumentation. Although the data are available in raw form for verification purposes, the data are not as yet in a suitable form for detailed analyses. Work continues on the data management aspects of the project and in continued monitoring of the test.

  2. Creep deformation of an unirradiated zircaloy nuclear fuel cladding tube under dry storage conditions

    NASA Astrophysics Data System (ADS)

    Mayuzumi, Masami; Onchi, Takeo

    1990-05-01

    Measurements of creep deformation were made on an internally gas pressurized tubular Zircaloy-4 specimen with plugs welded to its ends. Creep tests were conducted at temperatures between 577 and 693 K for holding times of up to 26640 ks, to formulate the creep equation needed for predicting creep strain during dry storage of spent fuel. Discussion was also given to the difference of creep behaviour between irradiated and unirradiated fuel cladding, indicating that the equation derived is applicable for predicting creep strain of spent fuel cladding during dry storage.

  3. Determination of research octane number of gasoline fuels by octane analyzer. Interim report

    SciTech Connect

    Chen, S.

    1983-04-01

    The Foxboro Laboratory Octane Analyzer was investigated as an improved, more reliable, and somewhat less complicated method for assessing octane quality. The Octane Analyzer's responses (induction time, peak area, and peak height) were correlated with the Research Octane Number (RON), the Motor Octane Number (MON), and the Antiknock Index (RON + MON/2) as determined by ASTM D 2699 and D2700 engine test methods. Among the three measured responses, peak height was found to give best correlation. In addition, the correlation was better with the RON and the Antiknock Index than it was with the MON. The Octane numbers of Gasohols and Coordinating Research Council (CRC) full-boiling range unleaded fuels did not correlate with the Analyzer's responses as well as did commercial unleaded gasoline fuels. In conclusion, the Octane Analyzer can be used as a screening test or as an alternate method for measuring the octane number of gasoline fuels.

  4. 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.

  5. Performance Spec. for Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shipping Port Spent Fuel Canisters

    SciTech Connect

    JOHNSON, D.M.

    2000-03-14

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders.

  6. 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

  7. Drying damaged K West fuel elements (Summary of whole element furnace runs 1 through 8)

    SciTech Connect

    LAWRENCE, L.A.

    1998-10-13

    N Reactor fuel elements stored in the Hanford K Basins were subjected to high temperatures and vacuum conditions to remove water. Results of the first series of whole element furnace tests i.e., Runs 1 through 8 were collected in this summary report. The report focuses on the six tests with breached fuel from the K West Basin which ranged from a simple fracture at the approximate mid-point to severe damage with cladding breaches at the top and bottom ends with axial breaches and fuel loss. Results of the tests are summarized and compared for moisture released during cold vacuum drying, moisture remaining after drying, effects of drying on the fuel element condition, and hydrogen and fission product release.

  8. Vehicle acceleration and fuel consumption when operated on JP-8 fuel. Interim report, 1 June 1988-28 February 1989

    SciTech Connect

    Owens, E.C.; Yost, D.M.; Lestz, S.J.

    1989-02-01

    A limited test program (eight combat and tactical vehicles) was conducted to obtain a quantitative estimate of the change in combat and tactical vehicle performance and fuel consumption that would occur when converting the military fleet to MIL-T-83133 JP-8 (F-34) fuel. Data specifically desired included startability and idle quality, acceleration rates, and fuel consumption. Also, a comparative assessment of the on-vehicle smoke production capabilities of combat vehicles with the two fuels was desired. As a result of these tests, it was determined that substitution of JP-8 for DF-2 reduced the acceleration rates, and thus power, of all vehicles tested except for the M928 and M1009 vehicles, which improved or remained the same. Also, all vehicles tested, except for the M88A1 light recovery vehicles, had fuel consumption increases with JP-8 that were at or below that predicted by the heating value difference between the two fuels. No drivability or idle problems occurred with any of the test vehicles.

  9. 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.

  10. Advanced analytical methodology for mobility fuels and lubricants applications. Interim report, Jun 90-Jun 91

    SciTech Connect

    Fodor, G.E.

    1991-06-01

    A review is given of state-of-the-art analytical chemical methodologies and instrumentation to provide timely evaluation of the quality of fuels and lubricants. An assessment is also provided for the possible use of these techniques under battlefield and near-battlefield conditions by marginally trained petroleum supply technician.

  11. Severe degradation of BWR fuel failures: Coolant activity analysis. Interim report

    SciTech Connect

    Klepfer, H.H.; Kjaer-Pedersen, N.; Rank, P.; Ozer, O.

    1993-11-01

    The Electric Power Research Institute (EPRI) has been investigating an apparent recent increase in the frequency of severely degraded failed fuel rods in BWRs. These incidents involve abnormal levels of coolant radioactivity from the failed rod, and are followed by a steep increase in the activity from tramp uranium, most of which is deposited on fuel rod surfaces and therefore tends to carry over to subsequent operating cycles. Analyses of coolant activity from several BWRs experiencing degradation of failed fuel rods were carried out with extensive use of the CHIRON code, developed for EPRI by S. Levy Incorporated. CHIRON is based on steady-state conditions, but its trending plot capability was extremely useful for studying the time variation of radiochemical parameters. It is observed that the failure degradation process usually starts three to eight months after a primary failure occurrence. It is possible from measurements of Np{sup 239} coolant activity to reach quite accurate estimates of the amount of fuel released from the failed rod(s) over time. The neptunium calculations can also be used to estimate the size/length of the cladding rupture. An equation has been developed to predict offgas at the start of the next cycle from offgas at end of cycle. It is concluded that a failed BWR fuel rod can deteriorate severely within a few months, and therefore all failure occurrences should be acted upon with the serious consequences of severe damage in mind. A set of recommended guidelines is established for monitoring and interpreting activity signals, and for acting upon failure observations.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

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

    PubMed

    Werle, Sebastian; Dudziak, Mariusz

    2014-06-17

    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. PMID:24938297

  17. 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.

  18. 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.

  19. 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.

  20. Estimates of Zircaloy integrity during dry storage of spent nuclear fuel: Final report

    SciTech Connect

    Miller, A.K.; Brooks, M.; Cheung, T.Y.; Tasooji, A.; Wood, J.C.; Kelm, J.R.; Surette, B.A.; Frost, C.R.

    1989-05-01

    The analytical and experimental work described in this report is intended to predict the integrity of light-water reactor (LWR) fuel rods when the fuel rods are stored dry. The analytical portion considered all failure mechanisms that could be expected to operate under dry storage conditions, including creep rupture, external oxidation stress-corrosion cracking (SCC), fatigue, and clad splitting by UO/sub 2/ oxidation. Existing physically based models were used to predict the probability that LWR fuel rod cladding will fail in 100 years, as a function of the temperature at which the rods are stored. In the experimental portion, SCC tests were conducted on irradiated Zircaloy cladding to determine characteristics under conditions relevant to dry storage. ''Precracked'' and ''smooth'' (with only small naturally occurring flaws) specimens of irradiated cladding were subjected to ''split ring'' tests at initial stresses on the order of the yield stress in a variety of atmospheres containing iodine or cesium/cadmium. Most precracked specimens failed by SCC, and about one-third of smooth specimens irradiated to fluence above 2.5 /times/ 10/sup 24/ n/m/sup 2/ also failed. However, the stresses present in these tests were much higher than those expected in stored fuel cladding; therefore, the experimental results do not necessarily indicate likely SCC problems in dry-storage fuel. 68 refs., 54 figs., 35 tabs.

  1. The Ontario Hydro dry irradiated fuel storage program and concrete integrated container demonstration

    SciTech Connect

    Armstrong, P.J.; Grande, L. )

    1990-05-01

    The practicality of loading irradiated fuel into a concrete cask underwater in an existing pool facility has been successfully demonstrated. The cask holds about 7.7 metric-tons-uranium. Special design features allow the cask to be used for dry storage, for transportation, and for disposal without re-handling the fuel. The cask, called the concrete integrated container, or CIC, has been developed. This paper describes the loading, monitoring, and IAEA-based transportation certification of testing of the CIC.

  2. 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.

  3. A combined wet/dry sipping cell for investigating failed TRIGA fuel elements

    SciTech Connect

    Hammer, J.; Gallhammer, H.; Bock, H.

    1988-07-01

    Investigation for a failed TRIGA fuel element is performed with the help of a combined wet/dry sipping cell, which has been designed and fabricated at the Atominstitut Vienna. In this sipping cell a TRIGA fuel element can be studied for fission product release, both at normal and at elevated temperatures. This report describes the design features of the sipping cell and the fission product identification procedure with the help of a high purity Germanium detector and a multichannel analyzer.

  4. Heavy fuel engine technology assessment. Interim report, August-December 1997

    SciTech Connect

    Palacios, C.F.; Owens, E.C.; Wood, C.D.

    1998-02-01

    The intent of this study was to survey existing state of the art heavy fuel (diesel) engine technology and recommend an approach to DoD for the acquisition of JP-8 capable engines for these applications. Equipment developers and item managers were surveyed to identity vehicles and equipment currently using gasoline engines, or situation in which engine limitations severely compromise developmental objectives. The characteristics of current state of the art diesel engine technology, along with what might be achievable for military applications, were then compared with these requirements to determine what engine approaches might satisfy the equipment needs. The final recommendation combines the following three steps to satisfy the requirements of the wide range of DoD engine applications: (1) Modify existing diesel engines to meet weight and power specifications to provide 10,000 DoD engines per year; (2) Design a new engine family utilizing commercial technology for most components to provide 33,000 DoD engines per year; and (3) Design an engine family of very high power density to provide 1,000 engines per year that can not be produced by the other two steps.

  5. 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

  6. 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.

  7. Combustion gas properties. Part 3: Hydrogen gas fuel and dry air

    NASA Technical Reports Server (NTRS)

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

    1985-01-01

    A series of computations has been made to produce the equilibrium temperature and gas composition for hydrogen 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 sample tables and figures are provided in this report.

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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)

  15. 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. PMID:17964144

  16. Process for producing dry, sulfur-free, CH[sub 4]-enriched synthesis or fuel gas

    SciTech Connect

    Child, E.T.; Lafferty, W.L. Jr.; Suggitt, R.M.; Jahnke, F.C.

    1993-08-03

    A process is described for the production of a dry, sulfur-free methane enriched synthesis gas or fuel gas stream comprising: (1) cooling a particulate-free raw synthesis or fuel gas feed stream comprising H[sub 2], CO, CO[sub 2], H[sub 2]O, N[sub 2], H[sub 2]S, COS and with or without methane to a temperature in the range of about 60 F to 130 F and separating out at least a portion of water condensate; (2) mixing together said cooled raw synthesis or fuel gas from (1) and a portion of cryogenic liquefied natural gas (LNG) thereby further cooling the new synthesis or fuel gas to a temperature in the range of about [minus]75 F to 60 F; (3) directly contacting the mixture from (2) in an acid-gas removal zone with liquid acid-gas absorbent solvent thereby absorbing sulfur-containing compounds, water, and at least a portion of the CO[sub 2], and thereby producing acid-gas rich liquid absorbent solvent containing dissolved water and a dry stream of methane enriched synthesis or fuel gas; (4) separating said acid-gas rich liquid absorbent from said dry stream of methane enriched synthesis or fuel gas comprising H[sub 2], CO, CH[sub 4], and substantially no sulfur-containing gas or moisture; (5) regenerating the separated acid-gas rich liquid absorbent solvent to remove the sulfur-containing gas and the dissolved water; and (6) introducing regenerated liquid acid-gas absorbent solvent into said acid gas removal zone.

  17. Thermal Analysis of Cold Vacuum Drying (CVD) of Spent Nuclear Fuel (SNF)

    SciTech Connect

    PIEPHO, M.G.

    2000-03-23

    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 N Reactor spent fuel. This analysis provides the basis for the MCO thermal behavior at the CVD Facility in support of the safety basis documentation.

  18. Dry and Wet Molecular Dynamics Simulations of Nafion® Polymer Electrolyte Fuel Cell Membrane

    NASA Astrophysics Data System (ADS)

    Yana, Janchai; Lee, Vannajan Sanghiran; Nimmanpipug, Piyarat; Dokmaisrijan, Supaporn; Aukkaravittayapun, Suparerk; Vilaithong, Thirapat

    The interactions between the hydronium ions and the waters in Nafion® polyelectrolyte membrane are relevant in the proton transfer process of fuel cell. To investigate a role of water in the proton transfer mechanism, molecular dynamic simulations have been performed for models of Nafion® side chains cluster with the water molecules and the hydronium ions comparing with dry system. After simulations, the trajectories were analyzed in term of intermolecular distances, potential energy, and radial distribution function.

  19. 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.

  20. 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.

  1. Hanford spent nuclear fuel cold vacuum drying process equipment skid modification work plan

    SciTech Connect

    Graves, D.B.

    1998-05-04

    This document provides the work plan for modifications to be made to the first article Process Equipment Skid for the Cold Vacuum Drying (CVD) process. The primary objective is to provide engineering configuration control for any modifications made to the Process Equipment Skid during proof of performance testing at the 306E Facility. Development Control procedures will be used to complete the design drawings and Procurement Specification W-441-Pl-FA. The Process Equipment Skid is a system for removing water and drying Spent Nuclear Fuel contained in Multi-Canister Overpacks. The skid contains the Vacuum Purge System and the Tempered Water System (VPS/TWS). The first article Process Equipment Skid, and subsequent production skids, will later be installed in the Cold Vacuum Drying Facility.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. Method for calculating the duration of vacuum drying of a metal-concrete container for spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Karyakin, Yu. E.; Nekhozhin, M. A.; Pletnev, A. A.

    2013-07-01

    A method for calculating the quantity of moisture in a metal-concrete container in the process of its charging with spent nuclear fuel is proposed. A computing method and results obtained by it for conservative estimation of the time of vacuum drying of a container charged with spent nuclear fuel by technologies with quantization and without quantization of the lower fuel element cluster are presented. It has been shown that the absence of quantization in loading spent fuel increases several times the time of vacuum drying of the metal-concrete container.

  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. Calculation of the process of vacuum drying of a metal-concrete container with spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Karyakin, Yu. E.; Lavrent'ev, S. A.; Pavlyukevich, N. V.; Pletnev, A. A.; Fedorovich, E. D.

    2012-01-01

    An algorithm and results of calculation of the process of vacuum drying of a metal-concrete container intended for long-term "dry" storage of spent nuclear fuel are presented. A calculated substantiation of the initial amount of moisture in the container is given.

  9. 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.

  10. 200 Area Interim Storage Area Technical Safety Requirements

    SciTech Connect

    CARRELL, R.D.

    2000-03-15

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

  11. 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.

  12. Determining leaf dry matter content using the normalized dry matter index and its possible application for estimating fuel moisture content

    Technology Transfer Automated Retrieval System (TEKTRAN)

    The Normalized Dry Matter Index (NDMI) was developed for the remote sensing of dry matter content using high-spectral resolution data. This narrow-band index is based on absorption at a C-H bond stretch overtone (1722 nm wavelength) and is correlated with dry matter content in fresh green leaves. ...

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. Automotive diesel-fuel filter-qualification methodology and preliminary screening results. Interim report, Jan 89-Sep 91

    SciTech Connect

    Bessee, G.B.; Westbrook, S.R.; Stavinoha, L.L.

    1992-01-01

    This report covers a program to develop a methodology to evaluate military vehicle fuel filters that would become part of a proposed military fuel filter specification. For this program, thirteen different fuel filters used on military and commercial vehicles were tested using a multipass fuel filter test stand. Each filter type was tested in triplicate. Test parameters measured included differential pressure across the filter, particulate contamination in both the influent and effluent fuel (measured gravimetrically), filter load capacity, and filter efficiency. The filter test results varied widely. Analysis of the results illustrated the need for better specification and control of filters used in Army fuel systems. The filtering media in some of the filters tended to separate or allow channeling at widely varying pressure drops. Some of the higher efficiency filters tested were also found to allow a significant number of large diameters particles to pass through.

  18. Comparison of diesel exhaust emissions using JP-8 and low-sulfur diesel fuel. Interim report, March 1994-March 1995

    SciTech Connect

    Yost, D.M.; Montalvo, D.A.

    1995-11-01

    Comparative emission measurements were made in two dynamometer-based diesel engines using protocol specified by the U.S. Environmental Protection Agency (EPA) and the California Air Resources Board (CARB). A single JP-8 fuel with a sulfur level of 0.06 wt% was adjusted to sulfur levels of 0.11 and 0.26 wt%. The emission characteristics of the three fuels were compared to the 1994 EPA certification low-sulfur diesel fuel (sulfur level equal to 0.035 wt%) in the Detroit Diesel Corporation (DDC) 1991 prototype Series 60 diesel engine and in the General Motors (GM) 6.2L diesel engine. Comparisons were made using the hot-start transient portion of the heavy-duty diesel engine Federal Test Procedure. Results from the Army study show that the gaseous emissions for the DDC Series 60 engine using kerosene-based JP-8 fuel are essentially equal to values obtained with the 0.035 wt% sulfur EPA certification diesel fuel, and that an approximate sulfur level of 0.21 wt% in kerosene-type JP-8 fuel would be equivalent to the 0.035 wt% sulfur reference fuel. Similarly, the regulated gaseous emissions for the GM 6.2L engine using JP-8 fuel are essentially equal to the values obtained with the 0.035 wt% sulfur EPA reference fuel. All sulfur levels of kerosene-type JP-8 fuel up to the 0.30 wt% MIL-T-83133 specification maximum would be equivalent to a 0.035 wt% sulfur EPA reference fuel.

  19. 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.

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

    SciTech Connect

    Durbin, Samuel G.; Morrow, Charles W.

    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.

  1. 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.

  2. 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.

  3. Diesel engine endurance tests using JP-8 fuel blended with used engine oil. Interim report November 1996--December 1997

    SciTech Connect

    Frame, E.A.; Yost, D.M.; Palacios, C.F.

    1998-07-01

    Tests were done to examine the feasibility of disposing of used engine oil from military vehicles by blending it with JP-8 engine fuel to be used in diesel vehicles. Two Army diesel engines were evaluated in cyclic endurance dynamometer test procedures using JP-8 fuel blended with 7.5% vol used oil. Results were compared to baseline performance using neat JP-8 fuel. The following major differences were observed when using blended fuel: Significant ashy deposits were found in the pre-combustion chamber of the 4-cycle diesel engine; indications of imminent exhaust valve burning (streaking) were found on the exhaust valves in the 2-cycle diesel engine. For both engines, condition was such that continuous use of 7.5 %vol blend would not be recommended. Considering it would take between 19--68 years for an Army engine to reach the end of endurance test condition, use of blended fuel 1 or 2 times per year is judged acceptable from an endurance standpoint.

  4. 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.

  5. 40 CFR 80.141 - Interim detergent gasoline program.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 17 2012-07-01 2012-07-01 false Interim detergent gasoline program. 80.141 Section 80.141 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.141 Interim detergent gasoline program. (a) Effective dates...

  6. 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 PAGESBeta

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

    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

  7. 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.

  8. Failure analysis of fuel-injection pumps from generator sets fueled with Jet A-1. Interim report, Nov 90-Jan 91

    SciTech Connect

    Lacey, P.I.; Lestz, S.J.

    1991-01-01

    The U.S. Department of Defense (DOD) has adopted the single fuel for the battlefield concept. Diesel fuel will be replaced by JP-8/Jet A-1 in compression ignition engines, thereby lowering the fuel logistics burden. These fuels have successfully undergone extensive testing in both the laboratory and in field trials. However, increased failure rates are being reported on a number of fuel-sensitive components during Operation Desert Shield in Saudi Arabia. Five failed Stanadyne rotary fuel injection pumps were returned to the Belvoir Fuels and Lubricants Research Facility (BFLRF) at Southwest Research Institute (SwRI) for disassembly and post-failure analysis. Particular attention was given to the possible effects of low-lubricity fuel. The results of the investigation indicate that most of the failures may be attributed to causes other than poor fuel lubricity. The reason for failure of specific components in two of the pumps could not be conclusively determines. However, it is believed that they would not have occurred as a result of short-term operation with Jet A-1.

  9. Potential benefits from the use of JP-8 fuel in military ground equipment. Interim report, October 1987-February 1989

    SciTech Connect

    Montemayor, A.F.; Stavinoha, L.L.; Lestz, S.J.; LePera, M.E.

    1989-02-01

    The U.S. DoD is moving toward the use of JP-8 as a single fuel for use in Europe. Many potential benefits are associated with exclusive use of JP-8 in U.S. DOD equipment. This study discusses these benefits and provides references for further study. Some of the benefits associated with use of JP-8 will be immediate, and some will require time to be appreciated. Some benefits will accrue during peacetime operations, and some will be most apparent during times of conflict. As JP-8 finds increasing use in field tests and conversions of military bases, there will no doubt be problems that arise alongside the benefits. Careful weighing of the benefits and problems will ultimately lead to optimal usage of fuel resources and hopefully, increased readiness. The main benefits associated with use of JP-8 in military ground equipment are simplified logistics, increased readiness, reduced exhaust emissions, and better lubricant life. The information contained in this report is intended to delineate those areas where the use of JP-8 will prove beneficial and alert those personnel that will be affected by the change. Introduction of JP-8 into the military system should proceed. The most useful demonstration programs will be operations that involve joint operations of forces to include Army ground and aviation activities. These operations should be monitored for benefits as well as possible problems and the lessons learned applied accordingly. Jet engine fuels; Kerosene; Fuel additives; Diesel fuels; Compression/ignition engines; Sulfur/exhaust emissions/particulates; Logistics; Military/ground/combat vehicles.

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

    SciTech Connect

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

    2014-01-01

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

  11. 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.

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

    Technology Transfer Automated Retrieval System (TEKTRAN)

    A cost model was developed for fuel ethanol production from barley based on the EDGE (Enhanced Dry Grind Enzymatic) process (Nghiem, et al., 2008). In this process, in addition to beta-glucanases, which is added to reduce the viscosity of the barley mash for efficient mixing, another enzyme, beta-...

  13. Wear analysis of diesel-engine fuel-injection pumps from military ground equipment fueled with Jet A-1. Interim report Jan-May 91

    SciTech Connect

    Lacey, P.I.

    1991-05-01

    The U.S. Department of Defense has adopted the single fuel for the battlefield concept. During Operation Desert Shield/Storm, Jet A-1 replaced diesel in many applications. A simultaneous increase in fuel injection pump failures was observed during that operation. Prior to its introduction, a number of studies had indicated that JP-8 is compatible with the current fleet of ground equipment. This report forms part of an ongoing study to define the fuel lubricity requirements of ground equipment. The report also details the wear and failure mechanisms observed from used pumps. The results indicate that, although Jet A-1 does increase wear, many other failure mechanisms are also prevalent.

  14. Tribe kills fuel rod proposal

    SciTech Connect

    1995-02-13

    This article is a review of nuclear utilities` efforts to find a repository of spent fuel rods. The rejection by the Mescalero Apaches of plans to build a waste repository on tribal lands has left a number of utilities scrambling to find interim solutions. Prairie Island will have to close before the end of the year unless a solution is found, and the Hope Creek/Salem units, exhausting there storage capacity within ten years, are considering dry-cask storage.

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. Spent nuclear fuel project, Cold Vacuum Drying Facility human factors engineering (HFE) analysis: Results and findings

    SciTech Connect

    Garvin, L.J.

    1998-07-17

    This report presents the background, methodology, and findings of a human factors engineering (HFE) analysis performed in May, 1998, of the Spent Nuclear Fuels (SNF) Project Cold Vacuum Drying Facility (CVDF), to support its Preliminary Safety Analysis Report (PSAR), in responding to the requirements of Department of Energy (DOE) Order 5480.23 (DOE 1992a) and drafted to DOE-STD-3009-94 format. This HFE analysis focused on general environment, physical and computer workstations, and handling devices involved in or directly supporting the technical operations of the facility. This report makes no attempt to interpret or evaluate the safety significance of the HFE analysis findings. The HFE findings presented in this report, along with the results of the CVDF PSAR Chapter 3, Hazards and Accident Analyses, provide the technical basis for preparing the CVDF PSAR Chapter 13, Human Factors Engineering, including interpretation and disposition of findings. The findings presented in this report allow the PSAR Chapter 13 to fully respond to HFE requirements established in DOE Order 5480.23. DOE 5480.23, Nuclear Safety Analysis Reports, Section 8b(3)(n) and Attachment 1, Section-M, require that HFE be analyzed in the PSAR for the adequacy of the current design and planned construction for internal and external communications, operational aids, instrumentation and controls, environmental factors such as heat, light, and noise and that an assessment of human performance under abnormal and emergency conditions be performed (DOE 1992a).

  20. 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

  1. 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

  2. Control of degradation of spent LWR (light-water reactor) fuel during dry storage in an inert atmosphere

    SciTech Connect

    Cunningham, M.E.; Simonen, E.P.; Allemann, R.T.; Levy, I.S.; Hazelton, R.F.

    1987-10-01

    Dry storage of Zircaloy-clad spent fuel in inert gas (referred to as inerted dry storage or IDS) is being developed as an alternative to water pool storage of spent fuel. The objectives of the activities described in this report are to identify potential Zircaloy degradation mechanisms and evaluate their applicability to cladding breach during IDS, develop models of the dominant Zircaloy degradation mechanisms, and recommend cladding temperature limits during IDS to control Zircaloy degradation. The principal potential Zircaloy cladding breach mechanisms during IDS have been identified as creep rupture, stress corrosion cracking (SCC), and delayed hydride cracking (DHC). Creep rupture is concluded to be the primary cladding breach mechanism during IDS. Deformation and fracture maps based on creep rupture were developed for Zircaloy. These maps were then used as the basis for developing spent fuel cladding temperature limits that would prevent cladding breach during a 40-year IDS period. The probability of cladding breach for spent fuel stored at the temperature limit is less than 0.5% per spent fuel rod. 52 refs., 7 figs., 1 tab.

  3. 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.

  4. 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.

  5. Time/motion observations of reactor loading, transportation, and dry unloading of an oversized truck spent-fuel shipment

    SciTech Connect

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

    1988-01-01

    This paper presents actual time/motion data for an oversize truck spent-fuel shipment from its origin, Surry, Virginia to its destination, Idaho National Engineering Laboratory (INEL). These data include the receipt of the empty cask at the reactor, wet-loading the cask, over-the-road or in-transit data, and receipt and dry unloading of the shipping cask at the receiving facility. Occupational doses were recorded at the Surry Power Plant as well as at INEL, and public doses were calculated for the in-transit dose analysis. This shipment was one of a series performed in support of a demonstration and evaluation of dry storage at INEL. The oversized shipment consisted of a TN-8L shipping cask loaded with three 10-yr-old pressurized water reactor assemblies. The total distance traveled was {approx}2800 miles, requiring 62 h including stops. The time required to receive and inspect the empty shipping cask and wet-load and release the shipment at the reactor was {approx}14.1 h, and the time to receive the loaded cask, dry-transfer the spent fuel to the storage cask, and release the empty cask and trailer at the INEL facility was {approx}8.2 h.

  6. 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.

  7. Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

    SciTech Connect

    Peacock, H.B. Jr.

    1999-10-21

    This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed.

  8. Remote sensing of fuel moisture content from the ratios of canopy water indices with a foliar dry matter index

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

    Fuel moisture content (FMC) is an important variable for predicting the occurrence and spread of wildfire. Foliar FMC was calculated as the ratio of leaf foliar water content (Cw) and dry matter content (Cm). Recently, the normalized dry matter index (NDMI) was developed for the remote sensing of Cm using high-spectral resolution data. This study explored the potential for remote sensing of FMC using the ratio of various vegetation water indices with NDMI. For leaf-scale simulations, all index ratios were significantly related to FMC. For canopy-scale simulations, ratio indices of the normalized difference infrared index (NDII) and normalized difference water index (NDWI) with NDMI predicted FMC with R2 values of 0.900 and 0.864, respectively. NDII/NDMI determined from leaf reflectance data had the highest correlation with FMC. Further investigation needs to be conducted to evaluate the effectiveness of this approach at canopy scales with airborne remote sensing data.

  9. 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.

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

    DOEpatents

    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.

  11. RADIOLYTIC AND THERMAL PROCESSES RELEVANT TO DRY STORAGE OF SPENT NUCLEAR FUELS

    EPA Science Inventory

    Thousands of tons of metallic uranium spent-nuclear-fuel (SNF) remain in water storage across the Department of Energy complex. For example, the Hanford Site K-Basins hold 2300 metric tons of spent fuel, much of it severely corroded. Similar situations exist elsewhere in the DOE ...

  12. 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…

  13. 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.

  14. Spent nuclear fuel project cold vacuum drying facility supporting data and calculation database

    SciTech Connect

    IRWIN, J.J.

    1999-02-26

    This document provides a database of supporting calculations for the Cold Vacuum Drying Facility (CVDF). The database 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-002, 1997, ''Cold Vacuum Drying Facility Design Requirements'', Rev. 2, and the CVDF Summary Design Report. The database contains calculation report entries for all process, safety and facility systems in the CVDF, a general CVD operations sequence and the CVDF System Design Descriptions (SDDs). This database has been developed for the SNFP CVDF Engineering 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.

  15. 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.

  16. 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.

  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.

    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.

  18. 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.

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

    DOEpatents

    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.

  20. FINAL REPORT (PART 1). RADIOLYTIC AND THERMAL PROCESSES RELEVANT TO DRY STORAGE OF SPENT NUCLEAR FUELS

    EPA Science Inventory

    The scientific and engineering demands of the Department of Energy (DOE) Environmental Restoration and Waste Management tasks are enormous. For example, several thousand metric tons of metallic uranium spent nuclear fuel (SNF) remain in water storage awaiting disposition. Of this...

  1. Development and calibration of the shielded measurement system for fissile contents measurements on irradiated nuclear fuel in dry storage.

    SciTech Connect

    Mosby, W. R.; Jensen, B. A.

    2002-05-31

    In recent years there has been a trend towards storage of Irradiated Nuclear Fuel (INF) in dry conditions rather than in underwater environments. At the same time, the Department of Energy (DOE) has begun encouraging custodians of INF to perform measurements on INF for which no recent fissile contents measurement data exists. INF, in the form of spent fuel from Experimental Breeder Reactor 2 (EBR-II), has been stored in close-fitting, dry underground storage locations at the Radioactive Scrap and Waste Facility (RSWF) at Argonne National Laboratory-West (ANL-W) for many years. In Fiscal Year 2000, funding was obtained from the DOE Office of Safeguards and Security Technology Development Program to develop and prepare for deployment a Shielded Measurement System (SMS) to perform fissile content measurements on INF stored in the RSWF. The SMS is equipped to lift an INF item out of its storage location, perform scanning neutron coincidence and high-resolution gamma-ray measurements, and restore the item to its storage location. The neutron and gamma-ray measurement results are compared to predictions based on isotope depletion and Monte Carlo neutral-particle transport models to provide confirmation of the accuracy of the models and hence of the fissile material contents of the item as calculated by the same models. This paper describes the SMS and discusses the results of the first calibration and validation measurements performed with the SMS.

  2. 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.

  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. PMID:24813011

  4. Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

    NASA Astrophysics Data System (ADS)

    Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

    2014-08-01

    Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

  5. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Master Equipment List

    SciTech Connect

    IRWIN, J.J.

    1999-09-21

    This document provides the master equipment list (MEL) for the Cold Vacuum Drying Facility (CVDF). The MEL was prepared to comply with DOE Standard 3024-98, Content of System Design Descriptions. The MEL 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 and the CVDF System Design Descriptions (SDD). The MEL identifies the SSCs and their safety functions, the design criteria, codes and standards, and quality assurance requirements that are required for establishing the safety design basis of the SSCs. The MEL also includes operating parameters, manufacturer information, and references the procurement specifications for the SSCs. This MEL shall be updated, expanded, and revised in accordance with future phases of the CVDF SAR, the SDD's, and CVDF operations.

  6. 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.

  7. 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. PMID:23837906

  8. 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.

  9. Coke-free dry reforming of model diesel fuel by a pulsed spark plasma at low temperatures using an exhaust gas recirculation (EGR) system

    NASA Astrophysics Data System (ADS)

    Sekine, Yasushi; Furukawa, Naotsugu; Matsukata, Masahiko; Kikuchi, Eiichi

    2011-07-01

    Dry reforming of diesel fuel, an endothermic reaction, is an attractive process for on-board hydrogen/syngas production to increase energy efficiency. For operating this dry reforming process in a vehicle, we can use the exhaust gas from an exhaust gas recirculation (EGR) system as a source of carbon dioxide. Catalytic dry reforming of heavy hydrocarbon is a very difficult reaction due to the high accumulation of carbon on the catalyst. Therefore, we attempted to use a non-equilibrium pulsed plasma for the dry reforming of model diesel fuel without a catalyst. We investigated dry reforming of model diesel fuel (n-dodecane) with a low-energy pulsed spark plasma, which is a kind of non-equilibrium plasma at a low temperature of 523 K. Through the reaction, we were able to obtain syngas (hydrogen and carbon monoxide) and a small amount of C2 hydrocarbon without coke formation at a ratio of CO2/Cfuel = 1.5 or higher. The reaction can be conducted at very low temperatures such as 523 K. Therefore, it is anticipated as a novel and effective process for on-board syngas production from diesel fuel using an EGR system.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. DATING: A computer code for determining allowable temperatures for dry storage of spent fuel in inert and nitrogen gases

    SciTech Connect

    Simonen, E.P.; Gilbert, E.R.

    1988-12-01

    The DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) code can be used to calculate allowable initial temperatures for dry storage of light-water-reactor spent fuel. The calculations are based on the life fraction rule using both measured data and mechanistic equations as reported by Chin et al. (1986). The code is written in FORTRAN and utilizes an efficient numerical integration method for rapid calculations on IBM-compatible personal computers. This report documents the technical basis for the DATING calculations, describes the computational method and code statements, and includes a user's guide with examples. The software for the DATING code is available through the National Energy Software Center operated by Argonne National Laboratory, Argonne, Illinois 60439. 5 refs., 8 figs., 5 tabs.

  15. 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.

  16. High temperature postirradiation materials performance of spent pressurized water reactor fuel rods under dry storage conditions

    SciTech Connect

    Einziger, R.E.; Atkin, S.D.; Pasupathi, V.; Stellrecht, D.E.

    1982-04-01

    Postirradiation studies on failure mechanisms of well-characterized pressurized water reactor rods were conducted for up to a year at 482, 510, and 571/sup 0/C in limited air and inert gas atmospheres. No cladding breaches occurred even though the tests operated many orders of magnitude longer in time than the lifetime predicted by Blackburn's analyses. The extended lifetime is due to significant creep strain of the Zircaloy cladding, which decreases the internal rod pressure. The cladding creep also contributes to radial cracks, through the external oxide and internal fuel-cladding chemical interaction layers, which propagated into and arrested in an oxygen stabilized alpha-Zircaloy layer. There were no signs of either additional cladding hydriding, stress corrosion cracking, or fuel pellet degradation. If irradiation hardening does not reduce the stress rupture properties of Zircaloy, a conservative maximum storage temperature of 400/sup 0/C based on a stress-rupture mechanism is recommended to ensure a 1000-yr cladding lifetime.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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…

  2. Two CdZnTe Detector-Equipped Gamma-ray Spectrometers for Attribute Measurements on Irradiated Nuclear Fuel

    SciTech Connect

    Hartwell, John Kelvin; Winston, Philip Lon; Marts, Donna Jeanne; Moore-McAteer, Lisa Dawn; Taylor, Steven Cheney

    2003-04-01

    Some United States Department of Energy-owned spent fuel elements from foreign research reactors (FRRs) are presently being shipped from the reactor location to the US for storage at the Idaho National Engineering and Environmental Laboratory (INEEL). Two cadmium zinc telluride detector-based gamma-ray spectrometers have been developed to confirm the irradiation status of these fuels. One spectrometer is configured to operate underwater in the spent fuel pool of the shipping location, while the other is configured to interrogate elements on receipt in the dry transfer cell at the INEEL’s Interim Fuel Storage Facility (IFSF). Both units have been operationally tested at the INEEL.

  3. 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…

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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 ...

  9. 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.

  10. 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.

  11. Deformation and fracture map methodology for predicting cladding behavior during dry storage

    SciTech Connect

    Chin, B.A.; Khan, M.A.; Tarn, J.C.L.

    1986-09-01

    The licensing of interim dry storage of light-water reactor spent fuel requires assurance that release limits of radioactive materials are not exceeded. The extent to which Zircaloy cladding can be relied upon as a barrier to prevent release of radioactive spent fuel and fission products depends upon its integrity. The internal pressure from helium and fission gases could become a source of hoop stress for creep rupture if pressures and temperatures were sufficiently high. Consequently, it is of interest to predict the condition of spent fuel cladding during interim storage for periods up to 40 years. To develop this prediction, deformation and fracture theories were used to develop maps. Where available, experimental deformation and fracture data were used to test the validity of the maps. Predictive equations were then developed and cumulative damage methodology was used to take credit for the declining temperature of spent fuel during storage. This methodology was then used to predict storage temperatures below which creep rupture would not be expected to occur except in fuel rods with pre-existing flaws. Predictions were also made and compared with results from tests conducted under abnormal conditions.

  12. Evaluation of hardness and wear resistance of interim restorative materials

    PubMed Central

    Savabi, Omid; Nejatidanesh, Farahnaz; Fathi, Mohamad Hossein; Navabi, Amir Arsalan; Savabi, Ghazal

    2013-01-01

    Background: The interim restorative materials should have certain mechanical properties to withstand in oral cavity. The aim of this study was to evaluate the hardness and wear resistance of interim restorative materials. Materials and Methods: Fifteen identical rectangular shape specimens with dimensions of 2 mm × 10 mm × 30 mm were made from 7 interim materials (TempSpan, Protemp 3 Garant, Revotek, Unifast LC, Tempron, Duralay, and Acropars). The Vickers hardness and abrasive wear of specimens were tested in dry conditions and after 1 week storage in artificial saliva. The depth of wear was measured using surface roughness inspection device. Data were subjected to Kruskal–Wallis and Mann–Whitney tests. The Pearson correlation coefficient was used to determine the relationship between hardness and wear (α =0.05). Results: TempSpan had the highest hardness. The wear resistance of TempSpan (in dry condition) and Revotek (after conditioning in artificial saliva) was significantly higher (P < 0.05). There was no statistically significant correlation between degree of wear and hardness of the materials (P = 0.281, r = −0.31). Conclusion: Hardness and wear resistance of interim resins are material related rather than category specified. PMID:23946734

  13. Status report on the spent fuel test-Climax, Nevada Test Site: A test of dry storage of spent fuel in a deep granite location

    SciTech Connect

    Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

    1982-12-31

    The Spent Fuel Test-Climax (SFT-C) is located at a depth of 420 m in the Climax granite at the Nevada Test Site. The test array contains 11 canistered PWR fuel assemblies, plus associated electrical simulators and electrical heaters. There are nearly 900 channels of thermal, radiation, stress, displacement, and test control instrumentation.

  14. 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.

  15. 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.

  16. 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.

  17. Development of a dry low-NOx gas turbine combustor for a natural-gas fueled 2MW co-generation system

    SciTech Connect

    Mori, Masaaki; Sato, Hiroshi

    1998-07-01

    A dry low-NOx gas turbine combustor has been developed for natural-gas fueled co-generation systems in the power range of 1--4MW. The combustor. called the Double Swirler Combustor, uses the lean premixed combustion to reduce NOx emission. The combustor is characterized by two staged lean premixed combustion with two coaxial annular burners and a simple fuel control system without the complex variable geometry. Substantially low NOx level has been achieved to meet the strict NOx regulation to co-generation systems in Japan. High combustion efficiency has been obtained for a wide operating range. In 1994, Tokyo Gas and Ishikawajima-Harima Heavy Industries initiated a collaborative program to develop a natural-gas fueled low NOx gas turbine engine for new 2MW class co-generation system, named IM270. The Double Swirler Combustor, originally developed by Tokyo Gas, was introduced into the natural gas fueled version of the IM270. Engine test of the first production unit was successfully conducted to confirm substantially low NOx level of less than 15 ppm (O{sub 2} = 16%) with the output power of more than 2MW. Test for the durability and the reliability of the system is being conducted at Tokyo Gas Negishi LNG Terminal in Kanagawa, Japan and successful results have been so far obtained.

  18. 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.

  19. 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.

  20. 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.

  1. Status report on the Spent-Fuel Test-Climax, Nevada Test Site: a test of dry storage of spent fuel in a deep granite location

    SciTech Connect

    Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

    1982-03-01

    The Spent Fuel Test-Climax (SFT-C) is located at a depth of 420 m in the Climax granite at the Nevada Test Site. The test array contains 11 canistered PWR fuel assemblies, plus associated electrical simulators and electrical heaters. There are nearly 900 channels of thermal, radiation, stress, displacement, and test control instrumentation. This paper is a general status report on the test, which started in May 1980.

  2. Spent fuel storage at Prairie Island: January 1995 status

    SciTech Connect

    Closs, J.; Kress, L.

    1995-12-31

    The disposal of spent nuclear fuel has been an issue for the US since the inception of the commercial nuclear power industry. In the past decade, it has become a critical factor in the continued operation of some nuclear power plants, including the two units at Prairie Island. As the struggles and litigation over storage alternatives wage on, spent fuel pools continue to fill and plants edge closer to premature shutdown. Due to the delays in the construction of a federal repository, many nuclear power plants have had to seek interim storage alternatives. In the case of Prairie Island, the safest and most feasible option is dry cask storage. This paper discusses the current status of the Independent Spent Fuel Storage Installation (ISFSI) Project at Prairie Island. It provides a historical background to the project, discusses the notable developments over the past year, and presents the projected plans of the Northern States Power Company (NSP) in regards to spent fuel storage.

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Central waste complex interim operational safety requirements

    SciTech Connect

    Bendixsen, R.B.; Ames, R.R., Fluor Daniel Hanford

    1997-03-20

    This Interim Operational Safety Requirements document supports the authorization basis for interim operations and identifies restrictions on interim operations for the disposal and storage of solid waste in the Central Waste Complex. The Central Waste Complex Interim Operational Safety Requirements provide the necessary controls on operations in the Central Waste Complex to ensure the radiological and hazardous material exposure will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, 1327 the public and the environment.