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Sample records for fuel transportation cask

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

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

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

  2. Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

    SciTech Connect

    Powell, F.P.

    1995-04-01

    This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask`s primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives.

  3. Parametric study of radiation dose rates from rail and truck spent fuel transport casks

    SciTech Connect

    Parks, C.V.; Hermann, O.W.; Knight, J.R.

    1985-08-01

    Neutron and gamma dose rates from typical rail and truck spent fuel transport casks are reported for a variety of spent PWR fuel sources and cask conditions. The IF 300 rail cask and NLI 1/2 truck cask were selected for use as appropriate cask models. All calculations (cross section preparation, generation of spent fuel source terms, radiation transport calculations, and dose evaluation) were performed using various modules of the SCALE computational system. Conditions or parameters for which there were variations between cases include: detector distance from cask, spent fuel cooling time, the setting of fuel or neutron shielding cavities to either wet or dry, the cobalt content of assembly materials, normal fuel assemblies and consolidated cannisters, the geometry mesh interval size, and the order of the angular quadrature set. 13 refs., 6 figs., 9 tabs.

  4. A spent fuel transportation cask for the twenty-first century

    SciTech Connect

    Nair, B.R.; Little, C.C.; Goedicke, F.E.

    1989-01-01

    An innovative transportation cask design for legal weight truck shipments of spent nuclear fuel is presented. The proposed approach presents a significant departure from conventional cask designs in that it uses titanium alloy, a material with a high strength-to-weight ratio which has no precedent in transportation cask certification. The significant increase in payload obtainable with the proposed approach, and the associated benefits such as reduced life cycle costs, lower personnel exposure, and lower transportation accident risks are discussed. Also included is the strategy for addressing the challenge of demonstration regulatory compliance for a transportation cask based on new technology. 8 refs., 4 figs., 1 tab.

  5. Routine methods for post-transportation accident recovery of spent fuel casks

    SciTech Connect

    Shappert, L.B.; Pope, R.B. ); Best, R.E. ); Jones, R.H. , Los Gatos, CA )

    1991-01-01

    Spent fuel casks and other large radioactive material packages have been examined to determine whether the designs are adequate to allow the casks to be recovered using conventional recovery methods following a transportation accident. Casks and similar packages are typically designed with, and handled by, trunnions that support the package during transport. These trunnions are considered the best cask feature with which to grapple the cask once it is no longer in its usual shipping mode. Following a transport accident, the trunnions may be buried or entangled so that they are not readily accessible to initiate the recovery process. To evaluate the effectiveness of applying traditional recovery methods to spent fuel casks, a workshop was held in which a series of accidents involving casks were postulated; the modes of transportation considered included truck, rail, and barge. These participants knowledgeable in transport, handling, and, in some cases, recovery of large, heavy containers attended. Participants concluded that the physical recovery of a cask involved in an accident, irrespective of where the accident occurs, would be a straightforward rigging operation and that the addition of specific recovery features (e.g., additional trunnions) to the cask appears unnecessary.

  6. Routine methods for post-transportation accident recovery of spent fuel casks

    SciTech Connect

    Shappert, L.B.; Pope, R.B.; Best, R.E.; Jones, R.H.

    1991-12-31

    Spent fuel casks and other large radioactive material packages have been examined to determine whether the designs are adequate to allow the casks to be recovered using conventional recovery methods following a transportation accident. Casks and similar packages are typically designed with, and handled by, trunnions that support the package during transport. These trunnions are considered the best cask feature with which to grapple the cask once it is no longer in its usual shipping mode. Following a transport accident, the trunnions may be buried or entangled so that they are not readily accessible to initiate the recovery process. To evaluate the effectiveness of applying traditional recovery methods to spent fuel casks, a workshop was held in which a series of accidents involving casks were postulated; the modes of transportation considered included truck, rail, and barge. These participants knowledgeable in transport, handling, and, in some cases, recovery of large, heavy containers attended. Participants concluded that the physical recovery of a cask involved in an accident, irrespective of where the accident occurs, would be a straightforward rigging operation and that the addition of specific recovery features (e.g., additional trunnions) to the cask appears unnecessary.

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

    SciTech Connect

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Duffey, T.A.; Sutherland, S.H.; Reardon, P.C.

    1992-11-01

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

  8. Certification challenges in the development of an innovative high payload capacity spent fuel transportation cask

    SciTech Connect

    Mair, B.R.; Severson, M.J.; Ciez, A.P. )

    1990-01-01

    The design approach and certification strategy used in the development of an innovative transportation cask for legal weight truck shipments of spent nuclear fuel is presented. The proposed approach represents a significant departure from conventional cask designs in that it uses titanium alloy, a material with a high strength-to-weight ratio which has no precedent in transportation cask certification. The significant increase in payload obtainable with the proposed approach, and the associated benefits such as reduced life cycle costs, lower personnel exposure, and lower transportation accident risks are discussed. 8 refs., 3 figs., 1 tab.

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

    SciTech Connect

    Not Available

    1980-05-01

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

  10. Study of minimum-weight highway transporters for spent nuclear fuel casks: Technical report

    SciTech Connect

    Hoess, J.A.; Drago, V.J.

    1989-05-01

    There are federal and state limits on the maximum tractor-trailer- payload combination and individual axle loads permissible on US highways. These can generally be considered as two sets, i.e., legal-weight and overweight limits. The number of individual shipments required will decrease as the capacity of the spent nuclear fuel cask increases. Thus, there is an incentive for identifying readily available minimum-weight tractors and trailers capable of safely and reliably transporting as large a cask as possible without exceeding the legal gross combination weight (GCW) of 80,000 lb or selected overweight GCW limit of 110,000 lb. This study identifies options for commercially available heavy-duty on-highway tractors and trailers for transporting proposed future loaded spent nuclear fuel casks. Loaded cask weights of 56,000 and 80,000 lb were selected as reference design points for the legal-weight and overweight transporters, respectively. The technical data on tractor and trailer characteristics obtained indicate that it is possible to develop a tractor-trailer combination, tailored for spent nuclear fuel transportation service, utilizing existing technology and commercially available components, capable of safely and reliably transporting 56,000 and 80,000-lb spent nuclear fuel casks without exceeding GCWs of 80,000 and 10,000 lb, respectively. 4 figs., 14 tabs.

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

    SciTech Connect

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

    2007-01-01

    On April 7, 1982, a tank truck and trailer carrying 8,800 gallons of gasoline was involved in an accident in the Caldecott tunnel on State Route 24 near Oakland, California. The tank trailer overturned and subsequently caught fire. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook analyses to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by truck. The Fire Dynamics Simulator (FDS) code developed by National Institute of Standards and Technology (NIST) was used to determine the thermal environment in the Caldecott tunnel during the fire. The FDS results were used to define boundary conditions for a thermal transient model of a truck transport cask containing spent nuclear fuel. The Nuclear Assurance Corporation (NAC) Legal Weight Truck (LWT) transportation cask was selected for this evaluation, as it represents a typical truck (over-the-road) cask, and can be used to transport a wide variety of spent nuclear fuels. Detailed analysis of the cask response to the fire was performed using the ANSYS® computer code to evaluate the thermal performance of the cask design in this fire scenario. This report describes the methods and approach used to assess the thermal response of the selected cask design to the conditions predicted in the Caldecott tunnel fire. The results of the analysis are presented in detail, with an evaluation of the cask response to the fire. The staff concluded that some components of smaller transportation casks resembling the NAC LWT, despite placement within an ISO container, could degrade significantly. Small transportation casks similar to the NAC LWT would probably experience failure of seals in this severe accident scenario. USNRC staff evaluated the radiological consequences of the cask response to the Caldecott tunnel fire. Although some

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

    SciTech Connect

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

    2004-10-03

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

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

    SciTech Connect

    Sanders, T.L. ); Jordan, H. . Rocky Flats Plant); Pasupathi, V. ); Mings, W.J. ); Reardon, P.C. )

    1991-09-01

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

  14. A Multi-function Cask for At-Reactor Storage of Short-Cooled Spent Fuel, Transport, and Disposal

    SciTech Connect

    Forsberg, C.W.

    2004-07-01

    The spent nuclear fuel (SNF) system in the United States was designed with the assumptions that SNF would be stored for several years in an at-reactor pool and then transported to reprocessing plants for recovery of fissile materials, that security would not be a major issue, and that the SNF burnups would be low. The system has evolved into a once-through fuel cycle with high-burnup SNF, long-term storage at the reactor sites, and major requirements for safeguards and security. An alternative system is proposed to better meet these current requirements. The SNF is placed in multi-function casks with the casks used for at-reactor storage, transport, and repository disposal. The cask is the handling package, provides radiation shielding, and protects the SNF against accidents and assault. SNF assemblies are handled only once to minimize accident risks, maximize security and safeguards by minimizing access to SNF, and reduce costs. To maximize physical protection, the cask body is constructed of a cermet (oxide particles embedded in steel, the same class of materials used in tank armor) and contains no cooling channels or other penetrations that allow access to the SNF. To minimize pool storage of SNF, the cask is designed to accept short-cooled SNF. To maximize the capability of the cask to reject decay heat and to limit SNF temperatures from short-cooled SNF, the cask uses (1) natural circulation of inert gas mixtures inside the cask to transfer heat from the SNF to the cask body and (2) an overpack with external natural-circulation, liquid-cooled fins to transfer heat from the cask body to the atmosphere. This approach utilizes the entire cask body area for heat transfer to maximize heat removal rates-without any penetrations through the cask body that would reduce the physical protection capabilities of the cask body. After the SNF has cooled, the cooling overpack is removed. At the repository, the cask is placed in a corrosion-resistant overpack before disposal

  15. FUEL CASK IMPACT LIMITER VULNERABILITIES

    SciTech Connect

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

    2009-02-09

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

  16. Fuel shipment experience, fuel movements from the BMI-1 transport cask

    SciTech Connect

    Bauer, Thomas L.; Krause, Michael G

    1986-07-01

    The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including several factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.

  17. Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

    SciTech Connect

    Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky; Miroslav Picek; Alexey Smirnov; Sergey Komarov; Edward Bradley; Alexander Dudchenko; Konstantin Golubkin

    2007-10-01

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing, testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.

  18. Status of spent-fuel shipping cask development

    SciTech Connect

    Hall, I.K.; Hinschberger, T.S.

    1989-01-01

    The purpose of the Cask Systems Development Program is to develop a variety of cask systems that can safely and economically transport commercial spent fuel and high-level waste from the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. This paper is limited to a discussion of the status of from-reactor spent-fuel cask development; future cask development plans include MRS-to-repository casks, specialty casks for nonstandard spent fuel and nonfuel materials, and defense high-level waste casks. Spent-fuel casks must be available in the late 1990s to support the U.S. Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) shipments from utilities. DOE-Idaho, with the support of EG G Idaho, Inc., Sandia National Laboratories, and selected cask developing contractors, has been assigned the responsibility for developing a new generation of cask systems. Four categories of spent fuel casks were initially proposed: (1) legal weight truck (LWT) casks (2) overweight truck (OWT) casks (3) rail/barge (R/B) casks (4) dual purpose (DP) storage/transport casks. Casks are being designed for reduced occupational radiation exposure at the receiving facility by facilitating the use of remote handling equipment. Automation of remote handling systems may be used to reduce cask turnaround time. Reducing turnaround time promotes reduced radiation exposure to occupational workers and improves cask utilization efficiency.

  19. Requirements of Cask-Storage and Cask-Transport Licensing According to IAEA 1996 Rules for On-Site Storage of Spent Fuel in Germany - From the Utility's Point of View

    SciTech Connect

    Schmidt, Paul; Schmidt, Konrad; Geiser, Heinz; Schmidt, Paul

    2007-07-01

    As agreed between the German government and the utilities in 2001, all spent fuel assemblies (FAs) are now stored inside spent-fuel casks in on-site storage buildings. With the cask being the only route of FA disposal, cask licensing has become vital not only for cask vendors, but also for the utilities. Recently, the German authorities have taken the switch from the IAEA 1985 to the IAEA 1996 rules as the basis of the cask transport license as an opportunity to open new areas of discussion. One example is the behavior of the fuel assemblies inside the cask during and after the most harmful design accidents. Radiation-induced embrittlement of high-burn-up fuel rods (55-65 GWd/tHM FA-avg.) may cause failure of fuel rods after the 9 m drop. If one conservatively assumes the leak-tight cask to be flooded with water and if one conservatively considers at the same time the fuel released from the high-burn-up rods to be fresh fuel, one can construct a situation with questionable criticality safety. Because of these and similar considerations, the cask licensing procedure in Germany has become difficult and slow. As a result, the German utilities can not reach the FA burn-up levels granted in their operating licenses, as there is no cask type licensed that would suffice. Cask vendors, fuel vendors and utilities must work together to solve this problem. (authors)

  20. Transportation capabilities of the existing cask fleet

    SciTech Connect

    Johnson, P.E.; Joy, D.S.; Wankerl, M.W.

    1991-01-01

    This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 46 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the Department of Energy (DOE) consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated. 5 refs., 4 tabs.

  1. BR-100 spent fuel shipping cask development

    SciTech Connect

    McGuinn, E.J.; Childress, P.C.

    1990-01-01

    Continued public acceptance of commercial nuclear power is contingent to a large degree on the US Department of Energy (DOE) establishing an integrated waste management system for spent nuclear fuel. As part of the from-reactor transportation segment of this system, the B W Fuel Company (BWFC) is under contract to the DOE to develop a spent-fuel cask that is compatible with both rail and barge modes of transportation. Innovative design approaches were the keys to achieving a cask design that maximizes payload capacity and cask performance. The result is the BR-100, a 100-ton rail/barge cask with a capacity of 21 PWR or 52 BWR ten-year cooled, intact fuel assemblies. 3 figs.

  2. Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

    SciTech Connect

    SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

    2000-07-01

    Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

  3. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    PubMed

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

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

    SciTech Connect

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-07-01

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

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

    SciTech Connect

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

    2008-10-01

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

  6. Transportation accident response of a high-capacity truck cask for spent fuel

    SciTech Connect

    O`Connell, W.J.; Glaser, R.E.; Johnson, G.L.; Perfect, S.A.; McGuinn, E.J.; Lake, W.H.

    1995-11-01

    Two of the primary goals of this study were (i) to check the structural and thermal performance of the GA-4 cask in a broad range of accidents and (ii) to carry out a severe-accidents analysis as had been addressed in the Modal Study but now using a specific recent cask design and using current-generation computer models and capabilities. At the same time, it was desired to compare the accident performance of the Ga-4 cask to that of the generic truck cask analyzed in the Modal Study. The same range of impact and fire accidents developed in the Modal Study was adopted for this study. The accident-description data base of the Modal Study categorizes accidents into types of collisions with mobile or fixed objects, non-collision accidents, and fires. The mechanical modes of damage may be via crushing, impact, or puncture. The fire occurrences in the Modal Study data are based on truck accident statistics. The fire types are taken to be pool fires of petroleum products from fuel tanks and/or cargoes.

  7. GNS spent fuel cask experience

    SciTech Connect

    Weh, R. )

    1993-05-01

    The Gesellschaft fuer Nuklear-Service mbH (GNS), which is owned by German utilities, is responsible for the management of spent fuel and nuclear waste on behalf of the German utilities operating nuclear power plants. This paper describes the spent reactor fuel and waste shipping and/or storage casks that GNS manufacturers for nuclear facilities in Germany, and worldwide. So far more than 30 different casks have been produced in quantities ranging from one to several hundred of each type. GNS participates in the German Support Program to assist the International Atomic Energy Agency (IAEA) in developing verification procedures for dry storage casks containing spent fuel. This activity is also summarized.

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

    SciTech Connect

    Dyer, R.S.; Barnes, E.; Snipes, R.L.; Guskov, V.; Makarchuk, T.

    2007-07-01

    Russia, stores large quantities of spent nuclear fuel (SNF) from submarine and ice-breaker nuclear powered naval vessels. This high-level radioactive material presents a significant threat to the Arctic and marine environments. Much of the SNF from decommissioned Russian nuclear submarines is stored either onboard the submarines or in floating storage vessels in Northwest and Far East Russia. Some of the SNF is damaged, stored in an unstable condition, or of a type that cannot currently be reprocessed. In many cases, the existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing all of this fuel from remote locations. Additional transport and storage options are required. Some of the existing storage facilities being used in Russia do not meet health and safety and physical security requirements. The U.S. has assisted Russia in the development of a new dual-purpose metal-concrete transport and storage cask (TUK-108/1) for their military SNF and assisted them in building several new facilities for off-loading submarine SNF and storing these TUK-108/1 casks. These efforts have reduced the technical, ecological, and security challenges for removal, handling, interim storage, and shipment of this submarine fuel. Currently, Russian licensing limits the storage period of the TUK-108/1 casks to no more than two years before the fuel must be shipped for reprocessing. In order to extend this licensed storage period, a system is required to condition the casks by removing residual water and creating an inert storage environment by backfilling the internal canisters with a noble gas such as argon. The U.S. has assisted Russia in the development of a mobile cask conditioning system for the TUK-108/1 cask. This new conditioning system allows the TUK 108/1 casks to be stored for up to five years after which the license may be considered for renewal for an additional five years or the fuel will be shipped to

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

    SciTech Connect

    Danese, F.L. ); Johnson, P.E.; Joy, D.S. )

    1990-01-01

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

  10. Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

    SciTech Connect

    Carbajo, J.J.; Lindner, C.N. )

    1992-01-01

    The design of spent-fuel casks for storage and transport requires modeling the cask for criticality, shielding, thermal, and structural analyses. While some parts of the cask are homogeneous, other regions are heterogeneous with different materials intermixed. For simplicity, some of the heterogeneous regions may be modeled as homogeneous. This paper evaluates the effect of homogenizing some regions of a cask on calculating radiation dose rates outside the cask. The dose rate calculations were performed with the one-dimensional discrete ordinates shielding XSDRNPM code coupled with the XSDOSE code and with the three-dimensional QAD-CGGP code. Dose rates were calculated radially at the midplane of the cask at two locations, cask surface and 2.3 m from the radial surface. The last location corresponds to a point 2 m from the lateral sides of a transport railroad car.

  11. Fire resistant nuclear fuel cask

    DOEpatents

    Heckman, Richard C.; Moss, Marvin

    1979-01-01

    The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked.

  12. 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.

  13. European experience in transport/storage cask for vitrified residues

    SciTech Connect

    Otton, Camille; Sicard, Damien

    2007-07-01

    Available in abstract form only. Full text of publication follows: Because of the evolution of burnup of spent fuel to be reprocessed, the high activity vitrified residues would not be transported in the existing cask designs. Therefore, TN International has decided in the late nineties to develop a brand new design of casks with optimized capacity able to store and transport the most active and hottest canisters: the TN{sup TM}81 casks currently in use in Switzerland and the TN{sup TM}85 cask which shall permit in the near future in Germany the storage and the transport of the most active vitrified residues defining a thermal power of 56 kW (kilowatts). The challenges for the TN{sup TM}81 and TN{sup TM}85 cask designs were that the geometry entry data were very restrictive and were combined with a fairly wide range set by the AREVA NC Specification relative to vitrified residue canister. The TN{sup TM}81 and the TN{sup TM}85 casks have been designed to fully anticipate shipment constraints of the present vitrified residue production. It also used the feedback of current shipments and the operational constraints and experience of receiving and shipping facilities. The casks had to fit as much as possible in the existing procedures for the already existing flasks such as the TN{sup TM}28 cask and TS 28 V cask, all along the logistics chain of loading, unloading, transport and maintenance. (authors)

  14. Used Fuel Cask Identification through Neutron Profile

    SciTech Connect

    Rauch, Eric Benton

    2015-11-20

    Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

  15. Near-term commercial spent fuel shipping cask requirements

    SciTech Connect

    Daling, P.M.

    1984-11-01

    This report describes an analysis of the near-term commercial light water reactor (LWR) spent fuel transportation system. The objective was to determine if the existing commercial spent fuel shipping cask fleet is adequate to provide the needed transportation services for the period of time the US government would be authorized to accept spent fuel for Federal Interim Storage (FIS). A spent fuel shipping cask supply-demand analysis was performed to evaluate the existing fleet size. The results of the shipping cask handling capability study indicated that by weight, 75% of the spent fuel shipments will be by truck (overweight plus legal-weight truck). From the results of the shipping cask supply-demand analysis it was concluded that, if utilities begin large-scale applications for FIS, the five legal-weight truck (LWT) casks currently in service would be inadequate to perform all of the needed shipments as early as 1987. This further assumes that a western site would be selected for the FIS facility. If the FIS site were to be located in the East, the need for additional LWT casks would be delayed by about two years. The overweight truck (OWT) cask fleet (two PWR and two BWR versions) will be adequate through 1992 if some shipments to FIS can be made several years before a reactor is projected to lose full core reserve. This is because OWT cask requirements increase gradually over the next several years. The feasibility of shipping before losing full core reserve has not been evaluated. Cask utilization requirements in later years will be reduced if some shipments can be made prior to the time they are actually needed. The existing three rail casks are adequate to perform near-term shipments. 18 references, 4 figures, 18 tables.

  16. Method for determining the fuel contribution to the source term in transport casks

    SciTech Connect

    Rashid, Y. R.; Lake, W. H.; Sanders, T. L.

    1989-01-01

    Detailed models and analytical procedures are applied to the many complex aspects of spent fuel in transport including characterization of the fuel's irradiation conditions and initial states at the time of shipment, determination of the dynamic forces on the fuel assemblies that result from regulatory transportation accidents, modeling and analysis of the fuel's mechanical response to these forces, and estimation of the releasable radioactivity in the event of cladding breach. The methodology adopts a combined deterministic/probabilistic analysis approach in which each aspect of the problem is appropriately treated on the basis of its level of determinability. The results are obtained in the form of failure probabilities for each regulatory event considered. 3 refs., 6 figs.

  17. Below Grade Assessment of Spent Nuclear Fuel (SNF) Cask Transport Route

    SciTech Connect

    CHENAULT, D.M.

    1999-11-23

    The report provides an assessment of the route for the SNF Fuel transport system from the K Basins to the CVDF and to the CSB. Results include the identification of any underground structures or utilities traveled over by the transport, the overburden depths for all locations identified, evaluation of the loading conditions, and determination of the effects of the loads on the structures and utilities.

  18. Development of the GA-4 and GA-9 legal weight spent fuel casks

    SciTech Connect

    Grenier, R.M.; Meyer, R.J.; Mings, W.J.

    1992-09-01

    GA is nearing the completion of the final design of two legal weight truck spent fuel shipping casks, the GA-4 Cask for PWR fuel and the GA-9 Cask for BWR fuel. GA is developing the casks under contract to the US Department of Energy (DOE) Field Office, Idaho, as part of the Office of Civilian Radioactive Waste Management (OCRWM) Cask Systems Development Program (CSDP). The casks will transport intact spent fuel assemblies fro commercial nuclear reactors sites to a monitored retrievable storage facility or a permanent repository. The DOE initiated the Cask Systems Development Program in response to the Nuclear Waste Policy Act of 1982 which made DOE responsible for managing the program for permanent disposal of spent nuclear fuel and high-level waste. This paper describes developmental and design verification testing programs, and the present status of the GA-4 and GA-9 Cask designs.

  19. Shielding analysis of the NAC-LWT cask with MTR fuel using SCALE

    SciTech Connect

    Napolitano, D.G.

    1995-12-31

    NAC International has used the SCALE Code Package extensively for transport and storage cask design. This includes the design of the NAC-STC dual purpose cask, the ENSA-DPT dual purpose cask as well as design modifications to the NAC-LWT cask. The NAC-LWT is a legal weight truck cask that was originally designed to transport one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Recently, this cask has been modified to transport up to 42 materials test reactor (MTR) fuel elements. This paper discusses the use of the SCALE package in performing a source term analysis of MTR fuel and shielding analysis of the NAC-LWT cask in support of a 10 CFR Part 71 license amendment for MTR fuel contents.

  20. ALARA studies on spent fuel and waste casks

    SciTech Connect

    Sutherland, S.H.

    1980-04-01

    In this report, some implications of applying the ALARA concept to cask designs for transporting spent fuel, high-level commercial and defense waste, and remote-handled transuranic waste are investigated. The XSDRNPM, one-dimensional radiation transport code, was used to obtain potential shield designs that would yield total dose rates at 1.8 m from the cask surface of 10, 5, and 2 mrem/h. Gamma shields of depleted uranium, lead, and steel were studied. The capacity of the casks was assumed to be 1, 4, or 7 elements or canisters, and the wastes were 1, 3, 5, and 10 years old. Depending on the dose rate, the cask empty weights and lifetime transportation costs were estimated.

  1. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect

    Forsberg, Charles W.; Dole, Leslie R.

    2007-07-01

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

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

    SciTech Connect

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

    1991-01-01

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

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

    SciTech Connect

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

    2001-11-20

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

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

    SciTech Connect

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

    1980-03-01

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

  5. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D. R.; Cuta, J. M.; Enderlin, C. W.

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  6. TRANSPORTATION CASK RECEIPT AND RETURN FACILITY WORKER DOSE ASSESSMENT

    SciTech Connect

    V. Arakali

    2005-02-24

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

  7. DYNAMIC NON LINEAR IMPACT ANALYSIS OF FUEL CASK CONTAINMENT VESSELS

    SciTech Connect

    Leduc, D

    2008-06-10

    Large fuel casks present challenges when evaluating their performance in the accident sequence specified in 10CFR 71. Testing is often limited because of cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing using simplified analytical methods. This paper details the use of dynamic non-linear analysis of large fuel casks using advanced computational techniques. Results from the dynamic analysis of two casks, the T-3 Spent Fuel Cask and the Hanford Un-irradiated Fuel Package are examined in detail. These analyses are used to fully evaluate containment vessel stresses and strains resulting from complex loads experienced by cask components during impacts. Importantly, these advanced analytical analyses are capable of examining stresses in key regions of the cask including the cask closure. This paper compares these advanced analytical results with the results of simplified cask analyses like those detailed in NUREG 3966.

  8. COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

    SciTech Connect

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Dodge, R.E.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS is a general thermal-hydraulic analysis computer code for prediction of material temperatures and fluid conditions in a wide variety of systems. The code has been validated for analysis of spent fuel storage systems, as part of the Commercial Spent Fuel Management Program of the US Department of Energy. The code solves finite volume equations representing the conservation equations for mass, moment, and energy for an incompressible single-phase heat transfer fluid. The fluid solution is coupled to a finite volume solution of the conduction equation in the solid structure of the system. This document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part 1 describes the conservation equations, constitutive models, and solution methods used in the code. Part 2 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part 3 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available.

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

    SciTech Connect

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

    2005-11-07

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

  10. Thermal Hydraulic Analysis of Spent Fuel Casks

    Energy Science and Technology Software Center (ESTSC)

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  11. Shipping and storage cask data for spent nuclear fuel

    SciTech Connect

    Johnson, E.R.; Notz, K.J.

    1988-11-01

    This document is a compilation of data on casks used for the storage and/or transport of commercially generated spent fuel in the US based on publicly available information. In using the information contained in the following data sheets, it should be understood that the data have been assembled from published information, which in some instances was not internally consistent. Moreover, it was sometimes necessary to calculate or infer the values of some attributes from available information. Nor was there always a uniform method of reporting the values of some attributes; for example, an outside surface dose of the loaded cask was sometimes reported to be the maximum acceptable by NRC, while in other cases the maximum actual dose rate expected was reported, and in still other cases the expected average dose rate was reported. A summary comparison of the principal attributes of storage and transportable storage casks is provided and a similar comparison for shipping casks is also shown. References to source data are provided on the individual data sheets for each cask.

  12. TN International and ITS operational feedback regarding the decommissioning of obsolete casks dedicated to the transport and/or storage of nuclear raw materials, fuel and used fuel

    SciTech Connect

    Blachet, L.; Bimet, F.; Rennesson, N.

    2008-07-01

    Within the AREVA group, TN International is a major actor regarding the design of casks and transportation for the nuclear cycle. In the early 2005, TN International has started the project of decommissioning some of its own equipment and was hence the first company ever in the AREVA Group to implement this new approach. In order to do so, TN International has based this project by taking into account the AREVA Sustainable Development Charter, the French regulatory framework, the ANDRA (Agence Nationale pour la Gestion des Dechets Radioactifs - National Agency for the radioactive waste management) requirements and has deployed a step by step methodology such as radiological characterization following a logical route. The aim was to define a standardized process with optimized solutions regarding the diversity of the cask's fleet. As a general matter, decommissioning of nuclear casks is a brand new field as the nuclear field is more familiar with the dismantling of nuclear facilities and/or nuclear power plant. Nevertheless existing workshops, maintenance facilities, measurements equipments and techniques have been exploited and adapted by TN International in order to turn an ambitious project into a permanent and cost-effective activity. The decommissioning of the nuclear casks implemented by TN International regarding its own needs and the French regulatory framework is formalized by several processes and is materialized for instance by the final disposal of casks as they are or in ISO container packed with cut-off casks and big bags filled with crushed internal cask equipments, etc. The first part of this paper aims to describe the history of the project that started with a specific environmental analysis which took into account the values of AREVA as regards the Sustainable Development principles that were at the time and are still a topic of current concern in the world. The second part will deal with the definition, the design and the implementation of the

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

    SciTech Connect

    Brisbin, S.A., Fluor Daniel Hanford

    1997-02-13

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

  14. 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

  15. The Performance of Spent Fuel Casks in Severe Tunnel Fires

    SciTech Connect

    Bajwa, C.S.; Easton, E.P.; Hansen, A.

    2006-07-01

    The Nuclear Regulatory Commission (NRC), working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of various spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. The thermal performance of three different spent fuel cask designs (HOLTEC HI-STAR 100, TransNuclear TN-68, and NAC-LWT) was evaluated with the ANSYS{sup R} and COBRA-SFS analysis codes, utilizing boundary conditions for the tunnel fire obtained using NIST's Fire Dynamics Simulator (FDS) code. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of these analyses are described in detail in Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario, NUREG/CR-6886, published in draft for comment in November 2005. Comments received by the NRC on NUREG/CR-6886 will be addressed in the final version of the report. (authors)

  16. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    SciTech Connect

    Diggs, J M

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables.

  17. 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.

  18. Assessment of spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    SciTech Connect

    Daling, P.M.

    1985-08-01

    Realistic truck/rail modal fractions are specifically needed to support the Monitored Retrievable Storage (MRS) and repository facility designs and envirionmental assessment activities. The objective of this study was to evaluate the spent fuel shipping cask handling capabilities at operating and planned commercial LWRs and use this information to estimate realistic truck/rail modal fractions. The cask handling parameter data collected in this study includes cask handling crane capabilities, dimensions of loading pools, structural limits, availability of rail service, past experience with spent fuel shipments (i.e., which cask was used.), and any other conditions which could impede or preclude use of a particular shipping cask. The results of this evaluation are presented for each reactor. A summary of the results which indicates the number of plants that are capable of handling each transport mode is presented. Note that two types of highway shipments are considered; legal-weight truck (LWT) and overweight truck (OWT). The primary differences between these two types of highway shipments are the size and cargo capacity of the spent fuel shipping casks. The OWT cask is roughly 50% heavier, 50% larger in diameter, and has a 300% larger cargo capacity. As a result of this size differential, some plants are capable of handling LWT casks but not OWT casks.

  19. Cosmic-ray imaging of spent fuel casks

    NASA Astrophysics Data System (ADS)

    Guardincerri, Elena; Durham, J. Matthew; Morris, Christopher; Poulson, Daniel; Plaud-Ramos, Kenie; Fabritius, Joseph; Bacon, Jeffrey; Winston, Philip; Chichester, David

    2015-10-01

    Muon radiography was used to image the inside of a partially loaded Westinghouse MC-10 dry cask containing spent nuclear fuel at Idaho National Laboratory. We present here the results of a 100 hours long measurement taken in May 2015 with two muon trackers placed outside the cask. The data clearly show the location of the missing fuel bundles and demonstrate the feasibility of using cosmic rays to monitor fuel casks against illicit diversion of their content.

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

    SciTech Connect

    Seager, K. D.; Gianoulakis, S. E.; Barrett, P. R.; Rashid, Y. R.; Reardon, P. C.

    1992-01-01

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

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

    SciTech Connect

    Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

    2002-07-01

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

  2. Release Fractions from Multi-Element Spent Fuel Casks Resulting from HEDD Attack

    SciTech Connect

    Luna, R. E.

    2006-07-01

    This paper provides a simple model for estimating the release of respirable aerosols resulting from an attack on a spent fuel cask using a high energy density device (HEDD). Two primary experiments have provided data on potential releases from spent fuel casks under HEDD attack. Sandia National Laboratories (SNL) conducted the first in the early 1980's and the second was sponsored by Gessellshaft fur Anlagen- and Reaktorsicherheit (GRS) in Germany and conducted in France in 1994. Both used surrogate spent fuel assemblies in real casks. The SNL experiments used un-pressurized fuel pin assemblies in a single element cask while the GRS tests used pressurized fuel pin assemblies in a 9-element cask. Data from the two test programs is reasonably consistent, given the differences in the experiments, but the use of the test data for prediction of releases resulting from HEDD attack requires a method for accounting for the effects of pin pressurization release and the ratio of pin plenum gas release to cask free volume (VR). To account for the effects of VR and to link the two data sources, a simple model has been developed that uses both the SNL data and the GRS data as well as recent test data on aerosols produced in experiments with single pellets subjected to HEDD effects conducted under the aegis of the International Consortium's Working Group on Sabotage of Transport and Storage Casks (WGSTSC). (authors)

  3. A cask fleet operations study

    SciTech Connect

    Not Available

    1988-03-01

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

  4. The development and testprogram of transport and storage casks for vitrified high level wastes

    SciTech Connect

    Spiker, H.; Hueggenberg, R.

    1992-12-31

    Reprocessing of irradiated fuel assemblies generates canisters filled with vitrified high level radioactive waste (HLW). The canisters are made of stainless steel and welded leak-tight. These HLW canisters are subject to transport from the fuel reprocessing plant to intermediate and final storage. Since 1983, a number of different packages based on the type B(U) concept of the IAEA were therefore designed, tested, licensed and manufactured for the transport and the interim storage of HLW canisters in flasks. The theoretical layout of the cask was tested by measurements on a prototype cask of the scale 1:1. The measured heat transfer characteristics of this newly developed transport and storage cask will be described. They can be written as a function Nu - C * Ra{sup m}, with the variable factor C, and the constant exponent m., The factor C is different for the vertical and the horizontal position of the cask.

  5. Evaluation of improvement potential for spent fuel cask handling

    SciTech Connect

    Franklin, A.L.

    1981-02-01

    This report describes the quantitative analysis of opportunities to improve the loading/unloading operations for spent fuel shipping casks. The improvement potential is defined as a reduction in the time for completion or worker exposure for the complete handling operations. Two casks have been chosen as representative of presently available shipping casks. These are the NAC-1/NFS-4 legal weight truck cask and the IF-300 rail cask. The handling operations for each of these casks are broken down into a series of sequential steps. The time for completion and worker exposure is described by a probability density function for each step. These step descriptions are then combined to form a base case description of the total loading/unloading operation. Potential improvement opportunities are evaluated by modifying the appropriate probability density function descriptors then recombining the steps to form a probabilistic description of the modified operation.

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

    SciTech Connect

    Dilger, Fred; Halstead, Robert J.; Ballard, James D.

    2012-07-01

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's

  7. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  8. 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.

  9. Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

    DOE PAGESBeta

    Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; Scaglione, John M.

    2016-06-15

    We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

  10. Integrated approach to trailer design for spent fuel casks

    SciTech Connect

    Osborne, D.M.; Burgoyne, R.M.; Grenier, R.M.; Meyer, R.J.

    1989-02-01

    General Atomics (GA) is developing the GA-4 and GA-9 spent fuel transportation systems. The scope of our contract includes spent fuel casks, legal weight trailers, and ancillary equipment. Recent structural failures of spent fuel trailers have focused attention on trailer design. As a major element of spent fuel transportation systems, the concerns address the adequacy of trailer performance requirements, structural design and analysis, and in-service inspection and maintenance procedures. In response to these concerns, GA has applied an integrated approach to the design of the GA-4 and GA-9 transportation systems. The objectives are to design reliable, high-integrity trailers and to demonstrate their performance by test. Once the design is complete, a prototype trailer will be fabricated and a performance test program conducted in accordance with a comprehensive test program. GA`s trailer test program will include both design and operations elements, and will be used to optimize the operations and maintenance plan. The results of this program will provide positive public and regulatory perception of trailer durability and will support the development of industry standards for both legal weight and overweight trailers for spent fuel applications. 2 figs.

  11. Integrated approach to trailer design for spent fuel casks

    SciTech Connect

    Osborne, D.M.; Burgoyne, R.M.; Grenier, R.M.; Meyer, R.J.

    1989-02-01

    General Atomics (GA) is developing the GA-4 and GA-9 spent fuel transportation systems. The scope of our contract includes spent fuel casks, legal weight trailers, and ancillary equipment. Recent structural failures of spent fuel trailers have focused attention on trailer design. As a major element of spent fuel transportation systems, the concerns address the adequacy of trailer performance requirements, structural design and analysis, and in-service inspection and maintenance procedures. In response to these concerns, GA has applied an integrated approach to the design of the GA-4 and GA-9 transportation systems. The objectives are to design reliable, high-integrity trailers and to demonstrate their performance by test. Once the design is complete, a prototype trailer will be fabricated and a performance test program conducted in accordance with a comprehensive test program. GA's trailer test program will include both design and operations elements, and will be used to optimize the operations and maintenance plan. The results of this program will provide positive public and regulatory perception of trailer durability and will support the development of industry standards for both legal weight and overweight trailers for spent fuel applications. 2 figs.

  12. 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.

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

    SciTech Connect

    Not Available

    1987-12-01

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

  14. SPENT FUEL CASK IMPACT LIMITER ATTACHMENT DESIGN DEFICIENCIES

    SciTech Connect

    Leduc, D; Jeffery England, J

    2007-10-16

    A recent structural analysis of the T-3 Spent Fuel Containment Cask found problems with the design of the attachment system. Assumptions in the original SARP concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. Similar weaknesses in the attachment system designs of other casks were also noted. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.

  15. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-24

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... issued a final rule on December 22,1994 (59 FR 65898), that approved the Standardized NUHOMS Cask System... 3, 1997 (62 FR 46517), this rule is classified as Compatibility ] Category ``NRC.'' Compatibility...

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

    SciTech Connect

    Hartman, D.J.; Miller, D.D.; Hill, R.R.

    1992-04-01

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

  17. Separator assembly for use in spent nuclear fuel shipping cask

    DOEpatents

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  18. 78 FR 8050 - Spent Fuel Cask Certificate of Compliance Format and Content

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-05

    ... COMMISSION 10 CFR Part 72 Spent Fuel Cask Certificate of Compliance Format and Content AGENCY: Nuclear... that governs the format and content of spent fuel storage cask Certificates of Compliance (CoCs... criteria for the format and content to be included in a spent fuel storage cask Certificate of...

  19. A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

    SciTech Connect

    Andress, D. and Associates, Inc., Kensington, MD ); Joy, D.S. ); McLeod, N.B. Associates, Inc., Oakton, VA ); Peterson, R.W. and Associates, Inc., Alexandria, VA ); Rahimi, M. )

    1991-01-01

    The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elements as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.

  20. Shipping Cask Studies with MOX Fuel

    SciTech Connect

    Pavlovichev, A.M.

    2001-05-17

    Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

  1. Status of spent-fuel transportation system development

    SciTech Connect

    Chapman, R.L.; Hall, I.K.

    1988-01-01

    The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems that can safely and economically be used to move commercial spent fuel and high-level waste from the generator to the federal repository or monitored retrievable storage facility. There are four initiatives to the CSDP, but only the first, from reactor casks, has been activated. This paper is limited to a discussion of the status of that initiative. Schedule objectives for the CSDP include development of spent-fuel cask systems by 1995 to support the Office of Civilian Radioactive Waste Management shipments of spent fuel from utilities beginning in the late 1990s. The US Department of Energy (DOE)-Idaho, with the support of EG G Idaho, Inc., Sandia National Laboratories, and selected cask development contractors, has been assigned the responsibility for developing a family of cask systems that are suitable for the task. Initially, four categories of spent-fuel casks were to be developed. They are legal-weight truck (LWT) casks, overweight truck (OWT) casks, rail/barge (R/B) casks, and dual purpose (DP) (storage/transport) casks. For a variety of reasons, OWT and DP cask development activities have been deferred. Program goals include developing a family of casks that will permit minimizing total system life cycle costs, ensure safety to the general public and to occupational workers, and attain public confidence in the transportation system.

  2. Automated shielding analysis sequences for spent fuel casks

    SciTech Connect

    Tang, J.S.; Parks, C.V.; Hermann, O.W.

    1987-01-01

    Two important Shielding Analysis Sequences (SAS) have recently been developed within the SCALE computational system. These sequences significantly enhance the existing SCALE system capabilities for evaluating radiation doses exterior to spent fuel casks. These new control module sequences (SAS1 and SAS4) and their capabilities are discussed and demonstrated, together with the existing SAS2 sequence that is used to generate radiation sources for spent fuel. Particular attention is given to the new SAS4 sequence which provides an automated scheme for generating and using biasing parameters in a subsequent Monte Carlo analysis of a cask.

  3. Fuel Element Transfer Cask Modelling Using MCNP Technique

    NASA Astrophysics Data System (ADS)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  4. Fuel Element Transfer Cask Modelling Using MCNP Technique

    SciTech Connect

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-05

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  5. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect

    Pope, R B; Diggs, J M

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  6. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    NASA Astrophysics Data System (ADS)

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

  7. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    SciTech Connect

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-12

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP)

  8. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2002-06-26

    Document presents analysis performed to demonstrate criticality safety of packaging spent PWR fuel assemblies currently located at the 324 Building into a NAC-1 cask. Interim storage of the cask is also documented.

  9. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    SciTech Connect

    Tang, J.S.

    1980-11-01

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm (12 in).

  10. Spent fuel integrity during transportation

    SciTech Connect

    Funk, C.W.; Jacobson, L.D.

    1980-01-01

    The conditions of recent shipments of light water reactor spent fuel were surveyed. The radioactivity level of cask coolant was examined in an attempt to find the effects of transportation on LWR fuel assemblies. Discussion included potential cladding integrity loss mechanisms, canning requirements, changes of radioactivity levels, and comparison of transportation in wet or dry media. Although integrity loss or degradation has not been identified, radioactivity levels usually increase during transportation, especially for leaking assemblies.

  11. Time and dose assessment of barge shipment and at-reactor handling of a CASTOR V/21 spent fuel storage cask

    SciTech Connect

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

    1992-04-01

    This report contains the results of a time/motion analysis and a radiation dose assessment made during the receipt from barge transport and the loading of CAst iron cask for Storage and Transport Of Radioactive material (CASTOR) V/21 storage casks with spent nuclear fuel at the Surry Power Station in Virginia during 1987. The study was a cooperative effort between Pacific Northwest Laboratory (PNL) and Virginia Electric and Power Company (Virginia Power), and was funded by the US Department of Energy (DOE) Transportation Program Office. In this study, cask handling activities were tracked at the Surry Power Station, tracing the transfer of the empty spent fuel storage cask from an ocean-going vessel to a barge for river transport through the activities required to place the loaded storage cask at an at-reactor storage location.

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

    SciTech Connect

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

    1986-02-01

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

  13. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-17

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8... the Holtec International HI-STORM 100 dry cask storage system listing within the ``List of Approved... other aspects of the HI-STORM 100 dry storage cask system. Because the NRC considers this...

  14. Transportation of spent MTR fuels

    SciTech Connect

    Raisonnier, D.

    1997-08-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

  15. Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport

    SciTech Connect

    Boshoven, J.K.

    1992-01-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the fresh fuel'' assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions.

  16. Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport

    SciTech Connect

    Boshoven, J.K.

    1992-08-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the ``fresh fuel`` assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions.

  17. TGC36 a dual purpose cask for the transport and interim storage of compacted waste (CSD-C)

    SciTech Connect

    Oldiges, O.; Boniface, J.M.

    2008-07-01

    According to contractual and international obligations, the German Utilities have to return the residues resulting from the reprocessing of nuclear fuel assemblies (compacted hulls and ends) to Germany. The new dual purpose cask TGC36 is a joint product from the two leading companies in the field development and manufactory of nuclear casks in Europe, GNS and TN International, is intended for the transport to the interim storage facility Ahaus and to be stored there for up several years. For the development and the delivery of the TGC36 cask, GNS and TN International formed the AGC Consortium based on German law to combine the special know how of both partners in the most efficient way. The design and the licensing strategy of the TGC36 are introduced in this paper. In conclusions: GNS and TNI have formed a consortium named AGC to design, license and manufacture an innovative cask for the transport and the interim storage of the compacted wastes resulting from the reprocessing of the German spent fuel. This cask has been optimized in order to offer a high capacity of loading, and allows a payload of 36 canisters, leading to a total mass of approximately 116 Mg in transport configuration. The success of this project requires a special effort from both partner companies, members of the consortium, and implies also an efficient management of simultaneous tasks during the licensing period and the manufacturing time of the first items of the cask. (authors)

  18. Intermodal transportation of spent fuel

    SciTech Connect

    Elder, H.K.

    1983-09-01

    Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate.

  19. 10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Conditions for spent fuel storage cask renewal. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks...

  20. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.236 Specific requirements for... storage cask, maximum heat designed to be dissipated, maximum spent fuel loading limit, condition of...

  1. 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

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

    SciTech Connect

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

    2006-04-01

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

  3. Quality assurance on DCI transport/storage casks for radioactive material

    SciTech Connect

    Bittner, R.; Ewest, E.; Weh, R.

    1993-12-31

    The Gesellschaft for Nuclear-Service (GNS), owned by German utilities, is responsible, on their behalf, for the management of spent fuel and radioactive waste, resulting from the operation of their nuclear power plants. The company`s other responsibilities include the development, design and fabrication of the transport and storage casks thus required. Questions of quality assurance obviously play an important part here. Using the example of a spent fuel cask of the CASTOR family, the performance of quality assurance measures is explained. These start with the company`s quality assurance system and include control activities during design and fabrication. In accordance with German regulations, fabrication and test sequences plans are used, which determine the inclusion of all parties involved at the time required. The interplay of different review authorities is being illustrated.

  4. Drop tests and numerical impact analyses of new cask designs for High Activity Waste (Haw) and spent fuel - updated BAM design testing experiences

    SciTech Connect

    Volzke, H.; Zencker, U.; Qiao, L.; Feutlinske, K.; Musolff, A.

    2007-07-01

    In Germany, several new cask designs by international vendors (Gesellschaft fuer Nuklear Service mbH (GNS), TN International (TNI), Mitsubishi Heavy Industries (MHI)) are under design testing and within official licensing procedures for transport and storage casks for spent fuel and high activity waste (HAW). BAM (the German Federal Institute for Materials Research and Testing) has been performing several extensive drop test series with prototype casks to evaluate the safety margins against mechanical test conditions. An important project is the new GNS cask design for HAW, the CASTOR{sup R} HAW 28M. Sixteen drop tests have been performed under transport conditions with a 1:2 scale cask model equipped with impact limiters and extensively instrumented with strain gauges and accelerometers. Additionally, the accident scenario inside a storage facility has been investigated by a cask drop without impact limiters onto a nearly unyielding target. This scenario is dominated by highly dynamic effects and interactions between the test object and the target. Complete safety assessments for such mechanical accident scenarios and highly loaded cask structures require additional numerical investigations. They are done by complex finite element (FE) calculations that provide detailed dynamic stress and strain analyses all over the cask structure and at such points where sensors can't be applied. In addition, differences between the material property quantities of the prototype cask and the minimum material property requirements for the cask series production can be investigated as well as dimensional tolerances. By example, the safety assessment method and some of its special aspects are illustrated by the cask drop without an impact limiter onto a hard foundation. The main aspects and challenges are to develop a sufficient computer model of the cask and foundation and to provide detailed interpretation of the large amount of measurement data for achieving good correlation

  5. 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. )

  6. 78 FR 73379 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-06

    ... issued a final rule (65 FR 25241; May 1, 2000) that approved the HI-STORM 100 Cask System design and... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... 3, 1997 (62 FR 46517), this direct final rule is classified as Compatibility Category...

  7. 76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-28

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI90 List of Approved Spent Fuel Storage Casks: HI... regulations to add the HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks... 13, 2011. SAR Submitted by: Holtec International, Inc. SAR Title: Safety Analysis Report on the...

  8. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    SciTech Connect

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs.

  9. Transportation accident scenarios for commercial spent fuel

    SciTech Connect

    Wilmot, E L

    1981-02-01

    A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

  10. 75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-07

    ...The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN), NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to Certificate of Compliance (CoC) Number 1030. Amendment No. 1 would modify the CoC to add Combustion Engineering 16x16 class fuel......

  11. 76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-13

    ...The U.S. Nuclear Regulatory Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN) NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to Certificate of Compliance (CoC) Number 1030. Amendment No. 1 would revise the definitions for Damaged Fuel......

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

    SciTech Connect

    1990-02-01

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

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

    SciTech Connect

    Koji Shirai

    2006-04-01

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

  14. Thermal evaluation of alternative shipping cask for irradiated experiments

    SciTech Connect

    Guillen, Donna Post

    2015-06-01

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

  15. Preparation, loading and storage of castor THTR/AVR-casks for spent fuel elements-part of the decommissioning of the high temperature reactor AVR

    SciTech Connect

    Theenhaus, R.; Halaszovich, S.; Storch, S.

    1994-12-31

    Shipment of spent fuel elements of the AVR reactor to the interim storage site at the Forschungszentrum Juelich KFA has started in August 1993. The handling procedure involves the preparation of the transport-storage casks, their remote loading with two stainless steel flasks, each of them containing 950 spherical fuel elements, the closing of the casks, the leak testing, the dose rate measurements, smear tests, the transportation and stacking of the casks and finally their connection with the permanent electronical leak control system. The handling strictly follows the manual which is part of the license and it is inspected by a member of the supervising ministry and an independent expert ordered by the ministry. Until the end of 1993 32,300 fuel elements have been transferred to the storage site.

  16. Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    SciTech Connect

    Daling, P.M.; Konzek, G.J.; Lezberg, A.J.; Votaw, E.F.; Collingham, M.I.

    1985-04-01

    This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%.

  17. A robotic system to conduct radiation and contamination surveys on nuclear waste transport casks

    SciTech Connect

    Harrigan, R.W.; Sanders, T.L.

    1990-06-01

    The feasibility of performing, numerous spent fuel cask operations using fully integrated robotic systems is under evaluation. Using existing technology, operational and descriptive software and hardware in the form of robotic end effectors are being designed in conjunction with interfacing cask components. A robotic radiation and contamination survey system has been developed and used on mock-up cask hardware to evaluate the impact of such fully automated operations on cask design features and productivity. Based on experience gained from the survey system, numerous health physics operations can be reliably performed with little human intervention using a fully automated system. Such operations can also significantly reduce time requirements for cask-receiving operations. 7 refs., 51 figs., 6 tabs.

  18. COBRA-SFS CYCLE 3. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  19. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel...

  20. THERMAL EVALUATION OF ALTERNATE SHIPPING CASK FOR GTRI EXPERIMENTS

    SciTech Connect

    Donna Post Guillen

    2014-06-01

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

  1. Spent fuel shipping cask handling capability assessment of 27 selected light water reactors

    SciTech Connect

    Konzek, G.J.; Daling, P.M.

    1984-11-01

    This report presents an assessment of the spent fuel shipping cask handling capabilities of those nuclear plants currently projected to lose full core reserve capability in their spent fuel storage basins in the near future. The purpose of this assessment is to determine which cask types, in the current fleet, each of the selected reactors can handle. The cask handling capability of a nuclear plant depends upon both external and internal conditions at the plant. The availability of a rail spur, the lifting capacity of the crane, the adequacy of clearances in the cask receiving, loading, and decontamination areas and similar factors can limit the types of casks that can be utilized at a particular plant. This report addresses the major facility capabilities used in assessing the types of spent fuel shipping casks that can be handled at each of the 27 selected nuclear plants approaching a critical storage situation. The results of this study cannot be considered to be final and are not intended to be used to force utilities to ship by a particular mode. In addition, many utilities have never shipped spent fuel. Readers are cautioned that the results of this study reflect the current situation at the selected plants and are based on operator perceptions and guidance from NRC related to the control of heavy loads at nuclear power plants. Thus, the cask handling capabilities essentially represent snap-shots in time and could be subject to change as plants further analyze their capabilities, even in the near-term. The results of this assessment indicate that 48% of the selected plants have rail access and 59% are judged to be candidates for overweight truck shipments (with 8 unknowns due to unavailability of verifiable data). Essentially all of the reactors can accommodate existing legal-weight truck casks. 12 references, 1 figure, 4 tables.

  2. Air Transport of Spent Nuclear Fuel (SNF) Assemblies

    SciTech Connect

    Haire, M.J.; Moses, S.D.; Shapovalov, V.I.; Morenko, A.

    2007-07-01

    Sometimes the only feasible means of shipping research reactor spent nuclear fuel (SNF) among countries is via air transport because of location or political conditions. The International Atomic Energy Agency (IAEA) has established a regulatory framework to certify air transport Type C casks. However, no such cask has been designed, built, tested, and certified. In lieu of an air transport cask, research reactor SNF has been transported using a Type B cask under an exemption with special arrangements for administrative and security controls. This work indicates that it may be feasible to transport commercial power reactor SNF assemblies via air, and that the cost is only about three times that of shipping it by railway. Optimization (i.e., reduction) of this cost factor has yet to be done. (authors)

  3. 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.

  4. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

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

    SciTech Connect

    1990-02-01

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

  6. Source storage and transfer cask: Users Guide

    SciTech Connect

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

    1985-04-01

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

  7. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND...

  8. 10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Conditions for spent fuel storage cask renewal. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  9. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  10. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Conditions for spent fuel storage cask reapproval. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  11. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND...

  12. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE...

  13. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Conditions for spent fuel storage cask reapproval. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  14. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  15. 10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Conditions for spent fuel storage cask renewal. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  16. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND...

  17. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false List of approved spent fuel storage casks. 72.214 Section 72.214 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent...

  18. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  19. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  20. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-14

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181... spent fuel storage cask designs. The NRC subsequently issued a final rule on November 21, 2008 (73 FR... 3, 1997 (62 FR 46517), this rule is classified as Compatibility Category ``NRC.'' Compatibility...

  1. The impact of using reduced capacity baskets on cask fleet size and cask fleet mix

    SciTech Connect

    Joy, D.S.; Johnson, P.E.; Andress, D.A.

    1993-06-01

    The Civilian Radioactive Waste Management System transportation system will encounter a wide range of spent fuel characteristics. Since the Initiative I casks are being designed to transport 10-year-old fuel with a burnup of 35,000 MWd/MTU, there is a good likelihood that a number of the cask shipments will need to be derated in order to meet the Nuclear Regulatory Commission radiation guidelines. This report discusses the impact of cask derating by using reduced-capacity baskets. Cask derating, while enhancing the ability to move spent fuel with a wider range of age and burnup characteristics, increases the number of shipments; the amount of equipment (cask bodies, baskets, etc.); and the number of visits to both shipping and receiving sites required to transport a specific amount of spent fuel.

  2. 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.

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

    SciTech Connect

    Branney, S.

    2011-07-01

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

  4. COBRA-SFS. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form the latest release of the code, Cycle 2.

  5. Application of surface complexation modeling to the understanding of transportation cask weeping

    SciTech Connect

    Granstaff, V.E.; Chambers, W.B.

    1993-11-01

    A new application for surface complexation modeling is described. These models, which describe chemical equilibria among aqueous and adsorbed species, have typically been used for predicting groundwater transport of contaminants by modeling the natural adsorbents as various metal oxides. We have shown that this type of modeling can also be used to explain stainless steel surface contamination and decontamination mechanisms. Stainless steel transportation casks that are submerged in a spent fuel storage pool at nuclear power stations, can become contaminated with radionuclides such as {sup 137}CS, {sup 134}Cs, and {sup 60}Co. Subsequent release or desorption of these contaminants under varying environmental conditions occasionally results in the phenomenon known as ``cask weeping.`` We have postulated that contaminants in the storage pool adsorb onto the hydrous metal oxide surface of the passivated stainless steel and are subsequently released during transportation, due to varying environmental factors, such as humidity, road salt, dirt, and acid rain. It is well known that 304 stainless steel has a chromium enriched passive surface layer; thus its adsorption behavior should be similar to that of chromium oxide. Presented here are adsorption data for Co{sup +2} on Cr{sup 2}O{sup 3} which simulate the stainless steel surface contamination. These data are interpreted using electrostatic surface complexation models. The FITEQL computer program was used to obtain the electrostatic model constants from the experimental data. Because the concentrations of contaminants in the storage pool are too low to be measured accurately by conventional chemical analysis techniques, MINTEQA2 can be used, with the fitted constants, to extrapolate the equilibria to the low concentrations representative of storage pool water.

  6. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    SciTech Connect

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

    2007-06-21

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

  7. Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

    SciTech Connect

    Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

    1986-12-01

    This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions.

  8. Transportation implications of a closed fuel cycle.

    SciTech Connect

    Bullard, Tim; Bays, Samuel; Dennis, Matthew L.; Weiner, Ruth F.; Sorenson, Ken Bryce; Dixon, Brent; Greiner, Miles

    2010-10-01

    Transportation for each step of a closed fuel cycle is analyzed in consideration of the availability of appropriate transportation infrastructure. The United States has both experience and certified casks for transportation that may be required by this cycle, except for the transport of fresh and used MOX fuel and fresh and used Advanced Burner Reactor (ABR) fuel. Packaging that had been used for other fuel with somewhat similar characteristics may be appropriate for these fuels, but would be inefficient. Therefore, the required neutron and gamma shielding, heat dissipation, and criticality were calculated for MOX and ABR fresh and spent fuel. Criticality would not be an issue, but the packaging design would need to balance neutron shielding and regulatory heat dissipation requirements.

  9. COBRA-SFS CYCLE 3: Code System for Thermal Hydraulic Analysis of Spent Fuel Casks

    Energy Science and Technology Software Center (ESTSC)

    2003-11-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  10. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    SciTech Connect

    Mickalonis, J. I.

    2015-08-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

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

    SciTech Connect

    Mickalonis, J.

    2014-06-01

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

  12. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    SciTech Connect

    Mickalonis, J. I.

    2015-08-31

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  13. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false List of approved spent fuel storage casks. 72.214 Section 72.214 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT.... Editorial Note: For Federal Register citations affecting § 72.214, see the List of CFR Sections...

  14. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false List of approved spent fuel storage casks. 72.214 Section 72.214 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT.... Editorial Note: For Federal Register citations affecting § 72.214, see the List of CFR Sections...

  15. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false List of approved spent fuel storage casks. 72.214 Section 72.214 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT... affecting § 72.214, see the List of CFR Sections Affected, which appears in the Finding Aids section of...

  16. 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.

  17. 77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-27

    ... confirmed for this direct final rule published November 14, 2011 at 76 FR 70331. ADDRESSES: You can access... (76 FR 70331), the NRC published a direct final rule amending its regulations at Title 10 of the Code... 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 2 AGENCY:...

  18. 77 FR 9515 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-17

    .... On May 1, 2000 (65 FR 25241), the NRC issued an amendment to 10 CFR part 72 that approved the HI... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... issued a final rule on May 1, 2000 (65 FR 25241), that approved the Holtec International cask design...

  19. 77 FR 24585 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-25

    ... February 17, 2012, at 77 FR 9515, is confirmed as May 2, 2012. ADDRESSES: Please refer to Docket ID NRC... . SUPPLEMENTARY INFORMATION: On February 17, 2012 (77 FR 9515), the NRC published a direct final rule amending its... 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY:...

  20. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-18

    ... License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also... subsequently issued a final rule on November 21, 2008 (73 FR 70587), that approved the MAGNASTOR Cask System..., 1997, and published in the Federal Register on September 3, 1997 (62 FR 46517), this rule is...

  1. 75 FR 24786 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-06

    ... provides text and image files of NRC's public documents. If you do not have access to ADAMS or if there are..., entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990... cask designs. The NRC subsequently issued a final rule on December 11, 2006 (71 FR 71463),...

  2. 75 FR 42292 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-21

    ... on the State. Plain Language The Presidential Memorandum, ``Plain Language in Government Writing... for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also... subsequently issued a final rule on March 9, 2000 (65 FR 12444), that approved the NAC-MPC cask design...

  3. Neutron measurements around storage casks containing spent fuel and vitrified high-level radioactive waste at ZWILAG.

    PubMed

    Buchillier, T; Aroua, A; Bochud, F O

    2007-01-01

    Spectrometric and dosimetric measurements were made around a cask containing spent fuel and a cask containing high-level radioactive waste at the Swiss intermediate waste and spent fuel storage facility. A Bonner sphere spectrometer, an LB 6411 neutron monitor and an Automess Szintomat 6134A were used to characterise the n-gamma fields at several locations around the two casks. The results of these measurements show that the neutron fluence spectra around the cask containing radioactive waste are harder and higher in intensity than those measured in the vicinity of the spent fuel cask. The ambient dose equivalents measured with the LB 6411 neutron monitor are in good agreement with those obtained using the Bonner spheres, except for locations with soft neutron spectra where the monitor overestimates the neutron ambient dose equivalent by almost 50%. PMID:17494980

  4. Structural design of concrete storage pads for spent-fuel casks

    SciTech Connect

    Rashid, Y.R.; Nickell, R.E.; James, R.J. )

    1993-04-01

    The loading experienced by spent fuel dry storage casks and storage pads due to potential drop or tip-over accidents is evaluated using state-of-the-art concrete structural analysis methodology. The purpose of this analysis is to provide simple design charts and formulas so that design adequacy of storage pads and dry storage casks can be demonstrated. The analysis covers a wide range of slab-design parameters, e.g., reinforcement ratio, slab thickness, concrete compressive strength, and sub-base soil compaction, as well as variations in drop orientation and drop height. The results are presented in the form of curves, giving the force on the cask as a function of storage pad hardness for various drop heights. In addition, force-displacement curves, deformed shapes, crack patterns, stresses and strains are given for various slab-design conditions and drop events. The utility of the results in design are illustrated through examples.

  5. Used Fuel Testing Transportation Model

    SciTech Connect

    Ross, Steven B.; Best, Ralph E.; Maheras, Steven J.; Jensen, Philip J.; England, Jeffery L.; LeDuc, Dan

    2014-09-24

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  6. Ageing of a neutron shielding used in transport/storage casks

    SciTech Connect

    Nizeyiman, Fidele; Alami, Aatif; Issard, Herve; Bellenger, Veronique

    2012-07-11

    In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

  7. Thermal evaluation of alternative shipping cask for irradiated experiments

    DOE PAGESBeta

    Guillen, Donna Post

    2015-06-01

    Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and post-irradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavitymore » of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. Furthermore, from a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.« less

  8. The Need for Confirmatory Experiments on the Radioactive Source Term from Potential Sabotage of Spent Nuclear Fuel Casks

    SciTech Connect

    PHILBIN, JEFFREY S.; HOOVER, MARK D.; NEWTON, GEORGE J.

    2002-04-01

    A technical review is presented of experiment activities and state of knowledge on air-borne, radiation source terms resulting from explosive sabotage attacks on spent reactor fuel subassemblies in shielded casks. Current assumptions about the behavior of irradiated fuel are largely based on a limited number of experimental results involving unirradiated, depleted uranium dioxide ''surrogate'' fuel. The behavior of irradiated nuclear fuel subjected to explosive conditions could be different from the behavior of the surrogate fuel, depending on the assumptions made by the evaluator. Available data indicate that these potential differences could result in errors, and possible orders-of-magnitude overestimates of aerosol dispersion and potential health effects from sabotage attacks. Furthermore, it is suggested that the current assumptions used in arriving at existing regulations for the transportation and storage of spent fuel in the U.S. are overly conservative. This, in turn, has led to potentially higher-than-needed operating expenses for those activities. A confirmatory experimental program is needed to develop a realistic correlation between source terms of irradiated fuel and unirradiated fuel. The motivations for performing the confirmatory experimental program are also presented.

  9. Comparative analyses of spent nuclear fuel transport modal options: Transport options under existing site constraints

    SciTech Connect

    Brentlinger, L.A.; Hofmann, P.L.; Peterson, R.W.

    1989-08-01

    The movement of nuclear waste can be accomplished by various transport modal options involving different types of vehicles, transport casks, transport routes, and intermediate intermodal transfer facilities. A series of systems studies are required to evaluate modal/intermodal spent fuel transportation options in a consistent fashion. This report provides total life-cycle cost and life-cycle dose estimates for a series of transport modal options under existing site constraints. 14 refs., 7 figs., 28 tabs.

  10. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    SciTech Connect

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years.

  11. Radiological source terms resulting from sabotage to transportation casks: Final report

    SciTech Connect

    Miller, N E; Fentiman, A W; Kuhlman, M R; Ebersole, H N; Trott, B D; Orban, J E

    1986-11-01

    The Nuclear Regulatory Commission (NRC) promulgated a rule, 10 CFR 73.37, which established requirements for safeguarding shipments of spent fuel to reduce the risk from acts of sabotage of highly radioactive materials. After the rule became effective, experimental programs conducted by Battelle for the NRC and by Sandia for the DOE showed the consequences of an attack using explosives on a shipment of PWR spent fuel were significantly less than had been indicated by earlier analytical studies. As a result, NRC is considering modifying the safeguards requirements. In support of NRC's efforts to modify the rule, Battelle has conducted additional experimental studies to evaluate the consequences of attacks on shipments of high-temperature gas-cooled reactor (HTGR) spent fuel, nonpower reactor (NPR) spent fuel, and vitrified high-level waste (HLW). Model casks containing surrogates of the spent fuels or high-level waste were penetrated by the jet from a precision shaped charge. Air samples collected after each test were used to estimate the quantities of respirable material released after the cask was penetrated. Results of the tests were scaled by specially developed scaling factors to estimate the releases that may occur from attacks on full-sized shipments of the materials. It was concluded that the sabotage of shipments of HTGR spent fuel, NPR spent fuel, or vitrified HLW should have no greater consequences than those predicted for shipments of PWR spent fuel.

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

    SciTech Connect

    Wu, T

    2009-06-25

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

  13. Code System to Calculate Radiation Dose Rates Relative to Spent Fuel Shipping Casks.

    Energy Science and Technology Software Center (ESTSC)

    1993-05-20

    Version 00 QBF calculates and plots in a short running time, three dimensional radiation dose rate distributions in the form of contour maps on specified planes resulting from cylindrical sources loaded into vehicles or ships. Shielding effects by steel walls and shielding material layers are taken into account in addition to the shadow effect among casks. This code system identifies the critical points on which to focus when designing the radiation shielding structure and wheremore » each of the spent fuel shipping casks should be stored. The code GRAPH reads the output data file of QBF and plots it using the HGX graphics library. QBF unifies the functions of the SMART and MANYCASK codes included in CCC-482.« less

  14. 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.

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

    SciTech Connect

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

    2014-05-30

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

  16. A new type-B cask design for transporting {sub 252}Cf

    SciTech Connect

    Simmons, C.M.

    2000-07-01

    A project to design, certify, and build a new US Department of Energy (DOE) Type B container for transporting >5 mg of {sup 252}Cf is more than halfway to completion. This project was necessitated by the fact that the existing Oak Ridge National Laboratory (ORNL) Type B containers were designed and built many years ago and thus do not have the records and supporting data that current regulations require. Once the new cask is available, it will replace the existing Type B containers. The cask design is driven by the unique properties of {sup 252}Cf, which is a very intense spontaneous fission neutron source and necessitates a large amount of neutron shielding. The cask is designed to contain up to 60 mg of {sup 252}Cf in the form of californium oxide or californium oxysulfate, in pellet, wire, or sintered material forms that are sealed inside small special-form capsules. The new cask will be capable of all modes of transport (land, sea, and air). The ORNL team, composed of technical and purchasing personnel and using rigorous selection criteria, chose NAC, International (NAC), as the subcontractor for the project. In January 1997, NAC started work on developing the conceptual design and performing the analyses. The original design concept was for a tungsten alloy gamma shield surrounded by two concentric shells of NS-4-FR neutron shield material. A visit to US Nuclear Regulatory Commission (NRC) regulators in November 1997 to present the conceptual design for their comments resulted in a design modification when the question of potential straight-line cracking in the NS-4-FR neutron shield material arose. NAC's modified design includes offset, wedgelike segments of the neutron shield material. The new geometry eliminates concerns about straight-line cracking but increases the weight of the packaging and makes the fabrication more complex. NAC has now completed the cask design and performed the analyses (shielding, structural, thermal, etc.) necessary to certify the

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

    SciTech Connect

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

    1996-09-01

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

  18. Followup audit of the cask development program

    SciTech Connect

    Not Available

    1994-03-15

    The Department of Energy is responsible for developing a system for the transportation and storage of spent nuclear fuel generated by utility companies. To carry out this responsibility, the Department of Energy established the Office of Civilian Radioactive Waste Management (Waste Management Office). The Waste Management office began development of a series of new shipping casks to transport the spent fuel. The purpose of this audit was to review the current development status of the cask designs; compare the original milestone dates to current milestone dates; and review the program funds that have been used to date on the development of these casks. The Office of Inspector General audited the cask development program in 1987. The audit report (DOE/IG-0244), recommended that program management establish minimum criteria that each cask must meet to qualify for further development funding. Our followup audit found that this recommendation had not been adequately implemented. As a result, the Waste Management office will spend an estimated $143 million on the cask development program and receive only two cask designs that were originally scheduled to cost $26 million. Moreover, it is not certain, at this time, whether those two cask designs will eventually receive the Nuclear Regulatory Commission certification. Historically, the program has experienced slippage in milestone dates and steady increases in total cost. Management generally agreed with our current recommendations to establish formal contingency plans to counter further delays, develop current baselines and schedules in sufficient detail to adequately control cask development schedules and costs, and reevaluate the current status of the casks under development for the purpose of justifying further development. Management has proposed actions to correct the milestone date slippages and continued growth in the total cost of the program.

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

    SciTech Connect

    Kee, A.T.

    1997-01-17

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

  20. Experimental data base for estimating the consequences from a hypothetical sabotage attack on a spent fuel shipping cask

    SciTech Connect

    Sandoval, R.P.; Luna, R.E.

    1986-01-01

    This paper describes the results of a program conducted at Sandia National Laboratories for the US Department of Energy to provide an experimental data base for estimating the radiological health effects that could result from the sabotage of a light water reactor spent fuel shipping cask. The primary objectives of the program were limited to: (1) evaluating the effectiveness of selected high energy devices (HED) in breaching full-scale spent fuel shipping casks, (2) quantifying and characterizing relevant aerosol and radiological properties of the released fuel, and (3) using the resulting experimental data to evaluate the radiological health effects resulting from a hypothetical attack on a spent fuel shipping cask in a densely populated urban area. 3 refs.

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

    SciTech Connect

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

    2010-07-18

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

  2. Cask fleet operations study

    SciTech Connect

    Not Available

    1988-01-01

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

  3. 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.

  4. 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.

  5. Transportation fuels from wood

    SciTech Connect

    Baker, E.G.; Elliott, D.C.; Stevens, D.J.

    1980-01-01

    The various methods of producing transportation fuels from wood are evaluated in this paper. These methods include direct liquefaction schemes such as hydrolysis/fermentation, pyrolysis, and thermochemical liquefaction. Indirect liquefaction techniques involve gasification followed by liquid fuels synthesis such as methanol synthesis or the Fischer-Tropsch synthesis. The cost of transportation fuels produced by the various methods are compared. In addition, three ongoing programs at Pacific Northwest Laboratory dealing with liquid fuels from wood are described.

  6. A comparison of spent fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    SciTech Connect

    Manson, S.J.; Gianoulakis, S.E.

    1994-04-01

    An examination of the effect of a realistic (though conservative) hot day environment on the thermal transient behavior of spent fuel shipping casks is made. These results are compared to those that develop under the prescribed normal thermal condition of 10 CFR 71. Of specific concern are the characteristics of propagating thermal waves, which are set up by diurnal variations of temperature and insolation in the outdoor environment. In order to arrive at a realistic approximation of these variations on a conservative hot day, actual temperature and insolation measurements have been obtained from the National Climatic Data Center (NCDC) for representatively hot and high heat flux days. Thus, the use of authentic meteorological data ensures the realistic approach sought. Further supporting the desired realism of the modeling effort is the use of realistic cask configurations in which multiple laminations of structural, shielding, and other materials are expected to attenuate the propagating thermal waves. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by enforcement of the regulatory environmental conditions of 10 CFR 71. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the prescribed regulatory conditions. However, the temperature differences are small enough that the normal conservative assumptions that are made in the course of typical cask evaluations should correct for any potential violations. The analysis demonstrates that diurnal temperature variations that penetrate the cask wall all have maxima substantially less than the corresponding regulatory solutions. Therefore it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the conditions of 10 CFR 71.

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

    SciTech Connect

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

    2011-01-14

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

  8. Discussion of Available Methods to Support Reviews of Spent Fuel Storage Installation Cask Drop Evaluations

    SciTech Connect

    Witte, M.

    2000-03-28

    Applicants seeking a Certificate of Compliance for an Independent Spent Fuel Storage Installation (ISFSI) cask must evaluate the consequences of a handling accident resulting in a drop or tip-over of the cask onto a concrete storage pad. As a result, analytical modeling approaches that might be used to evaluate the impact of cylindrical containers onto concrete pads are needed. One such approach, described and benchmarked in NUREG/CR-6608,{sup 1} consists of a dynamic finite element analysis using a concrete material model available in DYNA3D{sup 2} and in LS-DYNA,{sup 3} together with a method for post-processing the analysis results to calculate the deceleration of a solid steel billet when subjected to a drop or tip-over onto a concrete storage pad. The analysis approach described in NUREG/CR-6608 gives a good correlation of analysis and test results. The material model used for the concrete in the analyses in NUREG/CR-6608 is, however, somewhat troublesome to use, requiring a number of material constants which are difficult to obtain. Because of this a simpler approach, which adequately evaluates the impact of cylindrical containers onto concrete pads, is sought. Since finite element modeling of metals, and in particular carbon and stainless steel, is routinely and accurately accomplished with a number of finite element codes, the current task involves a literature search for and a discussion of available concrete models used in finite element codes. The goal is to find a balance between a concrete material model with a limited number of required material parameters which are readily obtainable, and a more complex model which is capable of accurately representing the complex behavior of the concrete storage pad under impact conditions. The purpose of this effort is to find the simplest possible way to analytically represent the storage cask deceleration during a cask tip-over or a cask drop onto a concrete storage pad. This report is divided into three sections

  9. COBRA-SFS thermal analysis of a sealed storage cask for the Monitored Retrievable Storage of spent fuel

    SciTech Connect

    Rector, D.R.; Wheeler, C.L.

    1986-01-01

    The COBRA-SFS (Spent Fuel Storage) computer code was used to predict temperature distributions in a concrete Sealed Storage Cask (SSC). This cask was designed for the Department of Energy in the Monitored Retrievable Storage (MRS) program for storage of spent fuel from commercial power operations. Analytical results were obtained for nominal operation of the SSC with spent fuel from 36 PWR fuel assemblies consolidated in 12 cylindrical canisters. Each canister generates 1650 W of thermal power. A parametric study was performed to assess the effects on cask thermal performance of thermal conductivity of the concrete, the fin material, and the amount of radial reinforcing steel bars (rebar). Seven different cases were modeled. The results of the COBRA-SFS analysis of the current cask design predict that the peak fuel cladding temperature in the SSC will not exceed the 37/sup 0/C design limit for the maximum spent fuel load of 19.8 kW and a maximum expected ambient temperature of 37.8/sup 0/C (100/sup 0/F). The results of the parametric analyses illustrate the importance of material selection and design optimization with regard to the SSC thermal performance.

  10. A preliminary investigation of the applicability of surface complexation modeling to the understanding of transportation cask weeping

    SciTech Connect

    Granstaff, V.E.; Chambers, W.B.; Doughty, D.H.

    1994-12-31

    A new application for surface complexation modeling is described. These models, which describe chemical equilibria among aqueous and adsorbed species, have typically been used for predicting groundwater transport of contaminants by modeling the natural adsorbents as various metal oxides. Our experiments suggest that this type of modeling can also explain stainless steel surface contamination and decontamination mechanisms. Stainless steel transportation casks, when submerged in a spent fuel storage pool at nuclear power stations, can become contaminated with radionuclides such as {sup 137}Cs, {sup 134}Cs, and {sup 60}Co. Subsequent release or desorption of these contaminants under varying environmental conditions occasionally results in the phenomenon known as {open_quotes}cask weeping.{close_quotes} We have postulated that contaminants in the storage pool adsorb onto the hydrous metal oxide surface of the passivated stainless steel and are subsequently released (by conversion from a fixed to a removable form) during transportation, due to varying environmental factors, such as humidity, road salt, dirt, and acid rain. It is well known that 304 stainless steel has a chromium enriched passive surface layer; thus its adsorption behavior should be similar to that of a mixed chromium/iron oxide. To help us interpret our studies of reversible binding of dissolved metals on stainless steel surfaces, we have studied the adsorption of Co{sup +2} on Cr{sub 2}O{sub 3}. The data are interpreted using electrostatic surface complexation models. The FITEQL computer program was used to obtain the model binding constants and site densities from the experimental data. The MINTEQA2 computer speciation model was used, with the fitted constants, in an attempt to validate this approach.

  11. 76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-08

    ... was published in the Federal Register on March 28, 2011 (76 FR 17019). This direct final rule amended....gov . SUPPLEMENTARY INFORMATION: On March 28, 2011 (76 FR 17019), the NRC published a direct final... 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY:...

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

    SciTech Connect

    Bucholz, J.A.

    1983-01-01

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

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

    SciTech Connect

    ARD, K.E.

    2000-04-19

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

  14. Fuel cell water transport

    DOEpatents

    Vanderborgh, Nicholas E.; Hedstrom, James C.

    1990-01-01

    The moisture content and temperature of hydrogen and oxygen gases is regulated throughout traverse of the gases in a fuel cell incorporating a solid polymer membrane. At least one of the gases traverses a first flow field adjacent the solid polymer membrane, where chemical reactions occur to generate an electrical current. A second flow field is located sequential with the first flow field and incorporates a membrane for effective water transport. A control fluid is then circulated adjacent the second membrane on the face opposite the fuel cell gas wherein moisture is either transported from the control fluid to humidify a fuel gas, e.g., hydrogen, or to the control fluid to prevent excess water buildup in the oxidizer gas, e.g., oxygen. Evaporation of water into the control gas and the control gas temperature act to control the fuel cell gas temperatures throughout the traverse of the fuel cell by the gases.

  15. Compton Dry-Cask Imaging System

    ScienceCinema

    None

    2013-05-28

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

  16. Compton Dry-Cask Imaging System

    SciTech Connect

    2011-01-01

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

  17. 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

  18. A validated methodology for evaluating burnup credit in spent fuel casks

    SciTech Connect

    Brady, M.C. ); Sanders, T.L. )

    1991-01-01

    The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in k{sub eff}. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs.

  19. 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.

  20. Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984

    SciTech Connect

    Not Available

    1986-04-01

    This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984.

  1. Evaluation of Cask Drop Criticality Issues at K Basin

    SciTech Connect

    GOLDMANN, L.H.

    2000-01-24

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

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

    SciTech Connect

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

    1987-06-01

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

  3. 78 FR 37927 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-25

    ... at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a new subpart L... rule on November 21, 2008 (73 FR 70587), that approved the MAGNASTOR Cask System design and added it to... Cask System design. The NRC published in the Federal Register on March 18, 2013 (78 FR 16601), a...

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

    SciTech Connect

    Mings, W.J. ); Koploy, M.A. )

    1992-01-01

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

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

    SciTech Connect

    Mings, W.J.; Koploy, M.A.

    1992-08-01

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

  6. CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect

    E.F. Loros

    2000-06-23

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

  7. 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.

  8. CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

    SciTech Connect

    Vinson, D.

    2010-07-11

    Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and at cask

  9. 75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-21

    ... closure ring for redundant closure into the Transportable Storage Canister (TSC) design; modification of the TSC and basket design to accommodate up to 68 La Crosse Boiling Water Reactor spent fuel... pattern)) that may contain undamaged Exxon fuel assemblies and damaged Exxon and Allis Chalmers...

  10. 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...

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

    SciTech Connect

    P. L. Winston

    2007-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2008-09-01

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

  13. Full-Scale Accident Testing in Support of Used Nuclear Fuel Transportation.

    SciTech Connect

    Durbin, Samuel G.; Lindgren, Eric R.; Rechard, Rob P.; Sorenson, Ken B.

    2014-09-01

    The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.

  14. PADD 5 Transportation Fuels Markets

    EIA Publications

    2015-01-01

    This study examines supply, demand, and distribution of transportation fuels in Petroleum Administration for Defense District (PADD) 5, a region that includes the western states of California, Arizona, Nevada, Oregon, Washington, Alaska, and Hawaii. For this study, transportation fuels include gasoline, diesel fuel, and jet fuel.

  15. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    SciTech Connect

    Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.; Gauld, Ian C.; Ilas, Germina; Marshall, William BJ J.

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

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

    SciTech Connect

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

    1986-06-01

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

  17. Estimated consequences from severe spent nuclear fuel transportation accidents

    SciTech Connect

    Arnish, J.J.; Monette, F.; LePoire, D.; Biwer, B.M.

    1996-06-01

    The RISKIND software package is used to estimate radiological consequences of severe accident scenarios involving the transportation of spent nuclear fuel. Radiological risks are estimated for both a collective population and a maximally exposed individual based on representative truck and rail cask designs described in the U.S. Nuclear Regulatory Commission (NRC) modal study. The estimate of collective population risk considers all possible environmental pathways, including acute and long-term exposures, and is presented in terms of the 50-y committed effective dose equivalent. Radiological risks to a maximally exposed individual from acute exposure are estimated and presented in terms of the first year and 50-y committed effective dose equivalent. Consequences are estimated for accidents occurring in rural and urban population areas. The modeled pathways include inhalation during initial passing of the radioactive cloud, external exposure from a reduction of the cask shielding, long-term external exposure. from ground deposition, and ingestion from contaminated food (rural only). The major pathways and contributing radionuclides are identified, and the effects of possible mitigative actions are discussed. The cask accident responses and the radionuclide release fractions are modeled as described in the NRC modal study. Estimates of severe accident probabilities are presented for both truck and rail modes of transport. The assumptions made in this study tend to be conservative; however, a set of multiplicative factors are identified that can be applied to estimate more realistic conditions.

  18. Development of NUPAC 140B 100 ton rail/barge cask

    SciTech Connect

    Not Available

    1990-04-01

    The NuPac 140-B 100 Ton Rail/Barge Shipping Cask Preliminary Design Report (PDR) presents a general introduction to, and description of, the NuPac 140-B Cask and its fuel payload. The NuPac 140-B Cask, Model: NuPac 140-B, is being designed by Nuclear Packaging, Inc., to meet or exceed all NRC and Department of Transportation regulations governing the shipment of radioactive material. Specifically the Cask is being developed as a safe means of transporting spent light-water-reactor (LWR) fuels from existing and proposed reactor facilities to a repository and/or a monitored retrievable storage (MRS) facility. The primary transportation mode is by railroad, although the shipping package is designed to be transported by barge and by truck shipment on a special overweight basis for short distances. This feature allows the servicing of reactor sites and other facilities which lack direct railroad access.

  19. Evaluation of the Cask Transportation Facility Modifications (CTFM) compliance to DOE order 6430.1A Project A.5 and A.6

    SciTech Connect

    ARD, K.E.

    2000-04-24

    This report was prepared to evaluate the compliance of CTFM to DOE Order 6430.1A. This document presents the results of an evaluation that was performed to assess compliance of the K West (KW) Cask Transportation Facility Modifications (CTFM) designs against applicable requirements of Department of Energy (DOE) Order 6430.1 A, General Design Criteria. This evaluation was grouped under two categories described as Cask Loadout System (CLS) and Cranes/Other Modifications.

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

    SciTech Connect

    BAZINET, G.D.

    2000-04-13

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

  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. Alcohol Transportation Fuels Demonstration Program

    SciTech Connect

    Kinoshita, C.M.

    1990-01-01

    Hawaii has abundant natural energy resources, especially biomass, that could be used to produce alternative fuels for ground transportation and electricity. This report summarizes activities performed during 1988 to June 1991 in the first phase of the Alcohol Transportation Fuels Demonstration Program. The Alcohol Transportation Fuels Demonstration Program was funded initially by the Energy Division of the State of Hawaii's Department of Business, Economic Development and Tourism, and then by the US Department of Energy. This program was intended to support the transition to an altemative transportation fuel, methanol, by demonstrating the use of methanol fuel and methanol-fueled vehicles, and solving the problems associated with that fuel. Specific objectives include surveying renewable energy resources and ground transportation in Hawaii; installing a model methanol fueling station; demonstrating a methanol-fueled fleet of (spark-ignition engine) vehicles; evaluating modification strategies for methanol-fueled diesel engines and fuel additives; and investigating the transition to methanol fueling. All major objectives of Phase I were met (survey of local renewable resources and ground transportation, installation of methanol refueling station, fleet demonstration, diesel engine modification and additive evaluation, and dissemination of information on alternative fueling), and some specific problems (e.g., relating to methanol fuel contamination during handling and refueling) were identified and solved. Several key issues emerging from Phase I (e.g., methanol corrosion, flame luminosity, and methanol-transition technoeconomics) were recommended as topics for follow-on research in subsequent phases of this program.

  3. COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.

    1986-12-01

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.

  4. Electricity as Transportation ``Fuel''

    NASA Astrophysics Data System (ADS)

    Tamor, Michael

    2013-04-01

    The personal automobile is a surprisingly efficient device, but its place in a sustainable transportation future hinges on its ability use a sustainable fuel. While electricity is widely expected to be such a ``fuel,'' the viability of electric vehicles rests on the validity of three assumptions. First, that the emissions from generation will be significantly lower than those from competing chemical fuels whether `renewable' or fossil. Second, that advances in battery technology will deliver adequate range and durability at an affordable cost. Third, that most customers will accept any functional limitations intrinsic to electrochemical energy storage. While the first two are subjects of active research and vigorous policy debate, the third is treated virtually as a given. Popular statements to the effect that ``because 70% of all daily travel is accomplished in less than 100 miles, mass deployment of 100 mile EVs will electrify 70% of all travel'' are based on collections of one-day travel reports such as the National Household Travel Survey, and so effectively ignore the complexities of individual needs. We have analyzed the day-to-day variations of individual vehicle usage in multiple regions and draw very different conclusions. Most significant is that limited EV range results in a level of inconvenience that is likely to be unacceptable to the vast majority of vehicle owners, and for those who would accept that inconvenience, battery costs must be absurdly low to achieve any economic payback. In contrast, the plug-in hybrid (PHEV) does not suffer range limitations and delivers economic payback for most users at realistic battery costs. More importantly, these findings appear to be universal in developed nations, with labor market population density being a powerful predictor of personal vehicle usage. This ``scalable city'' hypothesis may prove to a powerful predictor of the evolution of transportation in the large cities of the developing world.

  5. Development status of the GA-4 and GA-9 casks

    SciTech Connect

    Grenier, R.M.

    1992-01-01

    General Atomics (GA) has developed two legal-weight truck spent fuel shipping casks for transporting commercial reactor spent fuel. The GA-4 Cask carries four pressurized-water reactor (PWR) assemblies, and the GA-9 Cask carries nine boiling-water reactor (BWR) assemblies. Depleted uranium and a borated polymer are the gamma an neutron shielding materials. Type XM-19 stainless steel is the structural material used for the cask body, closure and the structure which supports the fuel assemblies. The impact limiters are made of aluminum honeycomb. Solid boron carbide, contained in the removable fuel support structure, provides poison for criticality control. The GA-4 Cask uses burnup credit to maintain criticality safety with spent fuel assemblies having enrichments greater than 3 wt% U-235. GA has conducted an extensive test program for the neutron shield material and the aluminum honeycomb impact limiters. Additional planned testing includes verification testing of a half-scale model to confirm the structural design, full-scale high and low temperature leak testing of the closure seal design, and endurance testing of the semitrailer design.

  6. Development status of the GA-4 and GA-9 casks

    SciTech Connect

    Grenier, R.M.

    1992-08-01

    General Atomics (GA) has developed two legal-weight truck spent fuel shipping casks for transporting commercial reactor spent fuel. The GA-4 Cask carries four pressurized-water reactor (PWR) assemblies, and the GA-9 Cask carries nine boiling-water reactor (BWR) assemblies. Depleted uranium and a borated polymer are the gamma an neutron shielding materials. Type XM-19 stainless steel is the structural material used for the cask body, closure and the structure which supports the fuel assemblies. The impact limiters are made of aluminum honeycomb. Solid boron carbide, contained in the removable fuel support structure, provides poison for criticality control. The GA-4 Cask uses burnup credit to maintain criticality safety with spent fuel assemblies having enrichments greater than 3 wt% U-235. GA has conducted an extensive test program for the neutron shield material and the aluminum honeycomb impact limiters. Additional planned testing includes verification testing of a half-scale model to confirm the structural design, full-scale high and low temperature leak testing of the closure seal design, and endurance testing of the semitrailer design.

  7. Criticality safety requirements for transporting EBR-II fuel bottles stored at INTEC

    SciTech Connect

    Lell, R. M.; Pope, C. L.

    2000-03-14

    Two carrier/shipping cask options are being developed to transport bottles of EBR-II fuel elements stored at INTEC. Some fuel bottles are intact, but some have developed leaks. Reactivity control requirements to maintain subcriticality during the hypothetical transport accident have been examined for both transport options for intact and leaking bottles. Poison rods, poison sleeves, and dummy filler bottles were considered; several possible poison materials and several possible dummy filler materials were studied. The minimum number of poison rods or dummy filler bottles has been determined for each carrier for transport of intact and leaking bottles.

  8. 75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-15

    ... on the State. Plain Language The Presidential Memorandum, ``Plain Language in Government Writing... FR 29181; July 18, 1990). This rule also established a new subpart L within 10 CFR part 72, entitled... FR 70587), that approved the MAGNASTOR cask design and added it to the list of NRC-approved...

  9. Analysis of radiation doses from operation of postulated commercial spent fuel transportation systems: Main report

    SciTech Connect

    Schneider, K.J.; Hostick, C.J.; Ross, W.A.; Peterson, R.W.; Smith, R.I.; Stiles, D.L.; Daling, P.M.; Weakley, S.A.; Grinde, R.B.; Young, J.R.

    1987-11-01

    This report contains a system study of estimated radiation doses to the public and workers resulting from the transport of spent fuel from commercial nuclear power reactors to a geologic repository. The report contains a detailed breakdown of activities and a description of time/distance/dose-rate estimates for each activity within the system. Collective doses are estimated for each of the major activities at the reactor site, in transit, and at the repository receiving facility. Annual individual doses to the maximally exposed individuals or groups of individuals are also estimated. A total of 17 alternatives and subalternatives to the postulated reference transportation system are identified, conceptualized, and their dose-reduction potentials and costs estimated. Resulting ratios of ..delta..cost/..delta..collective system dose for each alternative relative to the postulated reference transportation system are given. Most of the alternatives evaluated are estimated to provide both cost and dose reductions. Major reductions in transportation system dose and cost are estimated to result from using higher-capacity rail and truck casks, and particularly when replacing legalweight truck casks with ''advanced design'' overweight truck casks. The greatest annual dose reduction to the highest exposed individual workers (i.e., at the repository) is estimated to be achieved by using remote handling equipment for the cask handling operations at the repository. Additional shielding is also effective in reducing doses to both radiation workers at the reactor and repository and to transport workers. 69 refs., 36 figs., 156 tabs.

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

    SciTech Connect

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

    2010-08-11

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

  11. 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.

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

    SciTech Connect

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle; Wagner, John C.

    2013-07-01

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

  13. Analysis, scale modeling, and full-scale test of a railcar and spent-nuclear-fuel shipping cask in a high-velocity impact against a rigid barrier

    SciTech Connect

    Huerta, M.

    1981-06-01

    This report describes the mathematical analysis, the physical scale modeling, and a full-scale crash test of a railcar spent-nuclear-fuel shipping system. The mathematical analysis utilized a lumped-parameter model to predict the structural response of the railcar and the shipping cask. The physical scale modeling analysis consisted of two crash tests that used 1/8-scale models to assess railcar and shipping cask damage. The full-scale crash test, conducted with retired railcar equipment, was carefully monitored with onboard instrumentation and high-speed photography. Results of the mathematical and scale modeling analyses are compared with the full-scale test. 29 figures.

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

    SciTech Connect

    Not Available

    1988-01-01

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

  15. AREVA NP next generation fresh UO{sub 2} fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    SciTech Connect

    Doucet, M.; Landrieu, M.; Montgomery, R.; O' Donnell, B.

    2007-07-01

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO{sub 2} shipping casks being agreed within a lot of countries including USA, France, Germany, Belgium, Sweden, China, and South Africa - and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up-to-date shipping cask (so called MAP project) gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: - Preferential flooding; - Fuel rod lattice pitch expansion for full length of fuel assemblies; - Neutron absorber penalty; -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - Boral{sup TM} as neutron absorber; - High density polyethylene (HDPE) or Nylon as neutron moderator; - Foam as thermal and mechanical protection. The

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

    SciTech Connect

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

    1991-12-01

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

  17. Evaluation of impact limiter performance during end-on and slapdown drop tests of a one-third scale model storage/transport cask system

    SciTech Connect

    Yoshimura, H.R.; Bronowski, D.R.; Uncapher, W.L.; Attaway, S.W.; Bateman, V.I.; Carne, T.G.; Gregory, D.L. ); Huerta, M. )

    1990-12-01

    This report describes drop testing of a one-third scale model shipping cask system. Two casks were designed and fabricated by Transnuclear, Inc., to ship spent fuel from the former Nuclear Fuel Services West Valley reprocessing facility in New York to the Idaho National Engineering Laboratory for a long-term spent fuel dry storage demonstration project. As part of the NRC's regulatory certification process, one-third scale model tests were performed to obtain experimental data on impact limiter performance during impact testing. The objectives of the testing program were to (1) obtain deceleration and displacement information for the cask and impact limiter system, (2) obtain dynamic force-displacement data for the impact limiters, (3) verify the integrity of the impact limiter retention system, and (4) examine the crush behavior of the limiters. Two 30-ft (9-m) drop tests were conducted on a mass model of the cask body and scaled balsa and redwood-filled impact limiters. This report describes the results of both tests in terms of measured decelerations, posttest deformation measurements, and the general structural response of the system. 3 refs., 32 figs.

  18. Radiant heat transfer from storage casks to the environment

    SciTech Connect

    Carlson, R W; Hovingh, J; Thomas, G R

    1999-05-10

    A spent fuel storage cask must efficiently transfer the heat released by the fuel assemblies through the cask walls to the environment. This heat must be transferred through passive means, limiting the energy transfer mechanisms from the cask to natural convection and radiation heat transfer.. Natural convection is essentially independent of the characteristics of the array of casks, provided there is space between casks to permit a convection loop. Radiation heat transfer, however, depends on the geometric arrangement of the array of casks because the peripheral casks will shadow the interior casks and restrict radiant heat transfer from all casks to the environment. The shadowing of one cask by its neighbors is determined by a view factor that represents the fraction of radiant energy that leaves the surface of a cask and reaches the environment. This paper addresses the evaluation of the view factor between a centrally located spent fuel storage cask and the environment. By combining analytic expressions for the view factor of (1) infinitely long cylinders and (2) finite cylinders with a length-to-diameter ratio of 2 to represent spent fuel storage casks, the view factor can be evaluated for any practical array of spent fuel storage casks.

  19. Alternative Fuels in Transportation

    ERIC Educational Resources Information Center

    Kouroussis, Denis; Karimi, Shahram

    2006-01-01

    The realization of dwindling fossil fuel supplies and their adverse environmental impacts has accelerated research and development activities in the domain of renewable energy sources and technologies. Global energy demand is expected to rise during the next few decades, and the majority of today's energy is based on fossil fuels. Alternative…

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

    SciTech Connect

    Durbin, Samuel G.; Luna, Robert Earl

    2013-11-01

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

  1. Bioethanol: Fueling sustainable transportation

    SciTech Connect

    Neufeld, S.

    2000-05-25

    Ethanol made from biomass, or bioethanol, can positively impact the national energy security, the economy, and the environment. Producing and using bioethanol can help alleviate some of the negative impacts of the dependence on fossil fuels.

  2. Alternative transportation fuels

    SciTech Connect

    Askew, W. S.; McNamara, T. M.; Maxfield, D. P.

    1980-01-01

    The commercialization of alternative fuels is analyzed. Following a synopsis of US energy use, the concept of commercialization, the impacts of supply shortages and demand inelasticity upon commercialization, and the status of alternative fuels commercialization to date in the US are discussed. The US energy market is viewed as essentially numerous submarkets. The interrelationship among these submarkets precludes the need to commercialize for a specific fuel/use. However, the level of consumption, the projected growth in demand, and the inordinate dependence upon foreign fuels dictate that additional fuel supplies in general be brought to the US energy marketplace. Commercialization efforts encompass a range of measures designed to accelerate the arrival of technologies or products in the marketplace. As discussed in this paper, such a union of willing buyers and willing sellers requires that three general conditions be met: product quality comparable to existing products; price competitiveness; and adequate availability of supply. Product comparability presently appears to be the least problematic of these three requirements. Ethanol/gasoline and methanol/gasoline blends, for example, demonstrate the fact that alternative fuel technologies exist. Yet price and availability (i.e., production capacity) remain major obstacles. Given inelasticity (with respect to price) in the US and abroad, supply shortages - actual or contrived - generate upward price pressure and should make once-unattractive alternative fuels more price competitive. It is noted, however, that actual price competitiveness has been slow to occur and that even with price competitiveness, the lengthy time frame needed to achieve significant production capacity limits the near-term impact of alternative fuels.

  3. Status of spent fuel and radioactive waste management in Germany -- Development of casks and waste treatment equipment

    SciTech Connect

    Weh, R.

    1995-12-31

    At present, 20 nuclear power plants produce approximately one third of electricity in Germany. The proper disposal of the spent fuel and waste arising from operation of these plants is largely dealt with jointly by the German electricity suppliers, making use of their subsidiary company GNS. The utilities are responsible for the spent fuel and radioactive waste management; the final repositories are a federal matter. The German electricity suppliers thus concluded additional contracts with the reprocessing companies COGEMA in France and BNFL in the United Kingdom, amounting to approx. 5,500 t HM plus 2,800 t HM as new contracts. In the course of 1994, the German parliament passed a new act (Artikelgesetz) which provides the electricity suppliers an opportunity to choose between reprocessing and direct final disposal. This act is of major importance inasmuch as recent calculations show a clear financial advantage of direct final disposal. As these two disposal paths pose widely differing requirements to plant technology and storage capacities for spent fuel and waste, any kind of conceptual planing requires a high degree of flexibility. Direct final disposal of fuel assemblies as an alternative to reprocessing gains importance to the same degree as reprocessing is considered to be economically no longer justifiable. Limits to the permissible heat load in the planned Gorleben repository require long cooling periods for fuel assemblies which shall be achieved by intermediate storage in AFR stores. If the German utilities decide to stop reprocessing in favor of final disposal, the already constructed and licensed AFR stores at Gorleben (TBL-G) and Ahaus (BZA) will be used. Applications have been filed for the acceptance of fuel assemblies with a higher burn-up (> 60 MWd/t HM) or shorter cooling periods and of new casks with a higher capacity, e.g. the CASTOR V type series.

  4. The optimization of mechanical properties for nuclear transportation casks in ASTM A350 LF5

    SciTech Connect

    Price, S.; Honeyman, G.A.

    1997-12-31

    Transport flasks are required for the movement of spent nuclear fuel. Due to their nature of operation, it is necessary that these flasks are produced from forged steels with exceptional toughness properties. The material specification generally cited for flask manufacture is ASTM A350 Grade LF5 Class 1, a carbon-manganese-nickel alloy. The range of chemical analysis permitted by this specification is very broad and it is the responsibility of the material manufacturer to select a composition within this range which will satisfy all the mechanical properties requirements, and to ensure safe and reliable performance. Forgemasters Steel and Engineering Limited have experience in the manufacture of large high integrity fuel element flask forgings which extend over several decades. This experience and involvement in international standards in US, Europe and Japan has facilitated the development of an optimized analysis with a low carbon content, nickel levels towards the top end of the allowed range, a deliberate aluminum addition to control grain size and strictly controlled residual element levels. The resultant steel has excellent low temperature impact properties which greatly exceed the requirements of the specification. This analysis is now being adopted for the manufacture of all current transport flasks.

  5. Methods of producing transportation fuel

    DOEpatents

    Nair, Vijay; Roes, Augustinus Wilhelmus Maria; Cherrillo, Ralph Anthony; Bauldreay, Joanna M.

    2011-12-27

    Systems, methods, and heaters for treating a subsurface formation are described herein. At least one method for producing transportation fuel is described herein. The method for producing transportation fuel may include providing formation fluid having a boiling range distribution between -5.degree. C. and 350.degree. C. from a subsurface in situ heat treatment process to a subsurface treatment facility. A liquid stream may be separated from the formation fluid. The separated liquid stream may be hydrotreated and then distilled to produce a distilled stream having a boiling range distribution between 150.degree. C. and 350.degree. C. The distilled liquid stream may be combined with one or more additives to produce transportation fuel.

  6. Structural challenges in the development of a truck shipping cask for the OCRWM Cask Systems Development Program

    SciTech Connect

    Mello, R.M.; Severson, W.J.; Nair, B.R. . Nuclear Services Div.)

    1990-01-01

    The development of a spent fuel transportation cask design based on a structural material without licensing precedent presents many challenges. The US Nuclear Regulatory Commission (NRC) requires that any new material be qualified to meet the design and fabrication requirements of the ASME Boiler Pressure Vessel Code, Section III, Class 1. This paper discusses the strategy that is being implemented towards obtaining Code acceptance of a titanium alloy (3A1-2.5V). This alloy has been chosen as the principal structural material for a Legal Weight Truck cask being developed by Westinghouse for the US Department of Energy. The analysis approach used on some of the principal cask components is also presented. 5 refs., 8 figs., 3 tabs.

  7. Renewable and sustainable transportation fuels

    Technology Transfer Automated Retrieval System (TEKTRAN)

    This paper outlined the bioenergy research and production history as well as current status of renewable and sustainable transportation fuels in U.S. From starch based ethanol (first generation biofuel) to cellulosic ethanol (second generation biofuel), evident progress have been made in the past d...

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

    SciTech Connect

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

    2001-08-01

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

  9. 76 FR 70374 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-14

    ... Water Reactor and Boiling Water Reactor baskets and to replace the fuel tube orthogonal pitch with the minimum fuel tube outer diagonal dimension; Table 2.1-2, ``ASME Code Alternatives for MAGNASTOR components... SUPPLEMENTARY INFORMATION section of this document. You may submit comments by any one of the following...

  10. Safety evaluation for packaging (onsite) disposable solid waste cask

    SciTech Connect

    Flanagan, B.D., Westinghouse Hanford

    1996-12-20

    This safety evaluation for packaging (SEP) evaluates and documents the ability of the Disposable Solid Waste Cask (DSWC) to meet the packaging requirements of HNF-CM-2-14, Hazardous Material Packaging and Shipping, for the onsite transfer of special form, highway route controlled quantity, Type B fissile radioactive material. This SEP evaluates five shipments of DSWCs used for the transport and storage of Fast Flux Test Facility unirradiated fuel to the Plutonium Finishing Plant Protected Area.

  11. Alternatives to traditional transportation fuels: An overview

    SciTech Connect

    Not Available

    1994-06-01

    This report presents the first compilation by the Energy Information Administration (EIA) of information on alternatives to gasoline and diesel fuel. The purpose of the report is: (1) to provide background information on alternative transportation fuels and replacement fuels compared with gasoline and diesel fuel, and (2) to furnish preliminary estimates of alternative transportation fuels and alternative fueled vehicles as required by the Energy Policy Act of 1992 (EPACT), Title V, Section 503, ``Replacement Fuel Demand Estimates and Supply Information.`` Specifically, Section 503 requires the EIA to report annually on: (1) the number and type of alternative fueled vehicles in existence the previous year and expected to be in use the following year, (2) the geographic distribution of these vehicles, (3) the amounts and types of replacement fuels consumed, and (4) the greenhouse gas emissions likely to result from replacement fuel use. Alternative fueled vehicles are defined in this report as motorized vehicles licensed for on-road use, which may consume alternative transportation fuels. (Alternative fueled vehicles may use either an alternative transportation fuel or a replacement fuel.) The intended audience for the first section of this report includes the Secretary of Energy, the Congress, Federal and State agencies, the automobile manufacturing industry, the transportation fuel manufacturing and distribution industries, and the general public. The second section is designed primarily for persons desiring a more technical explanation of and background for the issues surrounding alternative transportation fuels.

  12. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2001-11-01

    This analysis references a previous analysis (Larson, 1999) for a qualitative acceptability argument, and an appropriate ANS standard (ANS, 1998), and a reference (Clark 1966) for safe cylinder diameters as a function of {sup 235}U enrichment in UO{sub 2} fuel rods (pins) in water, for a quantitative acceptability argument. The previous analysis established the criticality safety of PWR assemblies without broken pins, pin segments or powdered fuel. This addendum extends the previous range of applicability in accordance with the controlling procedure (FH 2001b). The operation, when conducted according to the established limits stated in this document, complies with the incredibility principle. The evaluations demonstrated criticality safety for PWR assemblies with broken pins and pin segments of UO{sub 2} fuel.

  13. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR... Fuel Storage Installation Docket Number: 72-1002 Certification Expiration Date: August 17, 2010 Model...: August 17, 2010 Model Number: NAC-C28 S/T Certificate Number: 1004. Initial Certificate Effective...

  14. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    SciTech Connect

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  15. Nondestructive Evaluation of the VSC-17 Cask

    SciTech Connect

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

    2006-01-01

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

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

    SciTech Connect

    Bracey, William; Bondre, Jayant; Shelton, Catherine; Edmonds, Robert

    2013-07-01

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

  17. Sensor system for fuel transport vehicle

    DOEpatents

    Earl, Dennis Duncan; McIntyre, Timothy J.; West, David L.

    2016-03-22

    An exemplary sensor system for a fuel transport vehicle can comprise a fuel marker sensor positioned between a fuel storage chamber of the vehicle and an access valve for the fuel storage chamber of the vehicle. The fuel marker sensor can be configured to measure one or more characteristics of one or more fuel markers present in the fuel adjacent the sensor, such as when the marked fuel is unloaded at a retail station. The one or more characteristics can comprise concentration and/or identity of the one or more fuel markers in the fuel. Based on the measured characteristics of the one or more fuel markers, the sensor system can identify the fuel and/or can determine whether the fuel has been adulterated after the marked fuel was last measured, such as when the marked fuel was loaded into the vehicle.

  18. Alcohol Transportation Fuels Demonstration Program. Phase 1

    SciTech Connect

    Kinoshita, C.M.

    1990-12-31

    Hawaii has abundant natural energy resources, especially biomass, that could be used to produce alternative fuels for ground transportation and electricity. This report summarizes activities performed during 1988 to June 1991 in the first phase of the Alcohol Transportation Fuels Demonstration Program. The Alcohol Transportation Fuels Demonstration Program was funded initially by the Energy Division of the State of Hawaii`s Department of Business, Economic Development and Tourism, and then by the US Department of Energy. This program was intended to support the transition to an altemative transportation fuel, methanol, by demonstrating the use of methanol fuel and methanol-fueled vehicles, and solving the problems associated with that fuel. Specific objectives include surveying renewable energy resources and ground transportation in Hawaii; installing a model methanol fueling station; demonstrating a methanol-fueled fleet of (spark-ignition engine) vehicles; evaluating modification strategies for methanol-fueled diesel engines and fuel additives; and investigating the transition to methanol fueling. All major objectives of Phase I were met (survey of local renewable resources and ground transportation, installation of methanol refueling station, fleet demonstration, diesel engine modification and additive evaluation, and dissemination of information on alternative fueling), and some specific problems (e.g., relating to methanol fuel contamination during handling and refueling) were identified and solved. Several key issues emerging from Phase I (e.g., methanol corrosion, flame luminosity, and methanol-transition technoeconomics) were recommended as topics for follow-on research in subsequent phases of this program.

  19. Methods of making transportation fuel

    DOEpatents

    Roes, Augustinus Wilhelmus Maria; Mo, Weijian; Muylle, Michel Serge Marie; Mandema, Remco Hugo; Nair, Vijay

    2012-04-10

    A method for producing alkylated hydrocarbons is disclosed. Formation fluid is produced from a subsurface in situ heat treatment process. The formation fluid is separated to produce a liquid stream and a first gas stream. The first gas stream includes olefins. The liquid stream is fractionated to produce at least a second gas stream including hydrocarbons having a carbon number of at least 3. The first gas stream and the second gas stream are introduced into an alkylation unit to produce alkylated hydrocarbons. At least a portion of the olefins in the first gas stream enhance alkylation. The alkylated hydrocarbons may be blended with one or more components to produce transportation fuel.

  20. Nuclear cask testing films misleading and misused

    SciTech Connect

    Audin, L. , Ossining, NY )

    1991-10-01

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

  1. Nuclear cask testing films misleading and misused

    SciTech Connect

    Audin, L.

    1991-10-01

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

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

    SciTech Connect

    Katsura Kajihara; Yasuhiro Aruga; Jun Shimojo; Hiroaki Taniuchi; Tsutomu Takeda; Masatosi Sasaki

    2002-07-01

    New enriched borated aluminum alloys manufactured by melting process are developed, which resulted in supplying structural basket materials for spent nuclear fuel packagings. In this process, the borated aluminum alloys were melted in a vacuum induction furnace at elevated temperature than that of ordinary aluminum melting processes. Boron dissolves into the matrix at the temperature of 1273 K or more, and fine aluminum diboride is precipitated and uniformly dispersed upon cooling rapidity. It is confirmed that boron is homogeneously dispersed with the fine particles of approximate 5 in average size in the product. Tensile strength and creep property at elevated temperature in 1 mass-%B 6061-T651 plate and 1 mass-%B 3004 extruded rectangular pipe as structural materials are examined. It is confirmed that the both of borated aluminum alloys have stable strength and creep properties that are similar to those of ordinary aluminum alloys. (authors)

  3. Radiological transportation risk assessment of the shipment of sodium-bonded fuel from the Fast Flux Test Facility to the Idaho National Engineering Laboratory

    SciTech Connect

    Green, J.R.

    1995-01-31

    This document was written in support of Environmental Assessment: Shutdown of the Fast Flux Test Facility (FFTF), Hanford Site, Richland, Washington. It analyzes the potential radiological risks associated with the transportation of sodium-bonded metal alloy and mixed carbide fuel from the FFTF on the Hanford Site in Washington State to the Idaho Engineering Laboratory in Idaho in the T-3 Cask. RADTRAN 4 is used for the analysis which addresses potential risk from normal transportation and hypothetical accident scenarios.

  4. 75 FR 57841 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-23

    ..., 2010, for the direct final rule that was published in the Federal Register on July 21, 2010 (75 FR....McCausland@nrc.gov . SUPPLEMENTARY INFORMATION: On July 21, 2010 (75 FR 42292), the NRC published a... Cask incorporating design features from the MAGNASTOR System for improved operability of the...

  5. FY 2012 USED FUEL DISPOSITION CAMPAIGN TRANSPORTATION TASK REPORT ON INL EFFORTS SUPPORTING THE MODERATOR EXCLUSION CONCEPT AND STANDARDIZED TRANSPORTATION

    SciTech Connect

    D. K. Morton

    2012-08-01

    Following the defunding of the Yucca Mountain Project, it is reasonable to assume that commercial used fuel will remain in storage for a longer time period than initially assumed. Previous transportation task work in FY 2011, under the Department of Energy’s Office of Nuclear Energy, Used Fuel Disposition Campaign, proposed an alternative for safely transporting used fuel regardless of the structural integrity of the used fuel, baskets, poisons, or storage canisters after an extended period of storage. This alternative assures criticality safety during transportation by implementing a concept that achieves moderator exclusion (no in-leakage of moderator into the used fuel cavity). By relying upon a component inside of the transportation cask that provides a watertight function, a strong argument can be made that moderator intrusion is not credible and should not be a required assumption for criticality evaluations during normal or hypothetical accident conditions of transportation. This Transportation Task report addresses the assigned FY 2012 work that supports the proposed moderator exclusion concept as well as a standardized transportation system. The two tasks assigned were to (1) promote the proposed moderator exclusion concept to both regulatory and nuclear industry audiences and (2) advance specific technical issues in order to improve American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3 rules for storage and transportation containments. The common point behind both of the assigned tasks is to provide more options that can be used to resolve current issues being debated regarding the future transportation of used fuel after extended storage.

  6. United States transportation fuel economics (1975 - 1995)

    NASA Technical Reports Server (NTRS)

    Alexander, A. D., III

    1975-01-01

    The United States transportation fuel economics in terms of fuel resources options, processing alternatives, and attendant economics for the period 1975 to 1995 are evaluated. The U.S. energy resource base is reviewed, portable fuel-processing alternatives are assessed, and selected future aircraft fuel options - JP fuel, liquid methane, and liquid hydrogen - are evaluated economically. Primary emphasis is placed on evaluating future aircraft fuel options and economics to provide guidance for future strategy of NASA in the development of aviation and air transportation research and technology.

  7. Fuel cell development for transportation: Catalyst development

    SciTech Connect

    Doddapaneni, N.

    1996-04-01

    Fuel cells are being considered as alternate power sources for transportation and stationary applications. With proton exchange membrane (PEM) fuel cells the fuel crossover to cathodes causes severe thermal management and cell voltage drop due to oxidation of fuel at the platinized cathodes. The main goal of this project was to design, synthesize, and evaluate stable and inexpensive transition metal macrocyclic catalysts for the reduction of oxygen and be electrochemically inert towards anode fuels such as hydrogen and methanol.

  8. As Built Verification Plan for Cask Transportation Facility Modifications (CTFM) Project A.5 and A.6

    SciTech Connect

    LANE, K.I.

    2000-04-20

    This document establishes an As-built Verification Plan (AVP) for implementing requirements in PHMC Engineering Requirements HNF-PRO-1819, Rev. 4, Sections 2.8.3.d and 2.10.8 and Spent Nuclear Fuels (SNF) Project Administrative Procedure EN-6-012-01. This AVP defines and implements approved processes to document the physical configuration of the project scope installed within the facility and identify discrepancies between the associated project engineering drawings and the field configuration, and the component index (CI) database as defined in AP EN 6-005-02. This AVP defines requirements for project activities verifying conformance of structures, systems, and components (SSCs) to project specified requirements.

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

    SciTech Connect

    SMITH, R.J.

    1998-11-17

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

  10. Alternatives to traditional transportation fuels 1996

    SciTech Connect

    1997-12-01

    Interest in alternative transportation fuels (ATF`s) has increased in recent years due to the drives for cleaner air and less dependence upon foreign oil. This report, Alternatives to Traditional Transportation Fuels 1996, provides information on ATFs, as well as the vehicles that consume them.

  11. Transportation fuels for the 21st century

    EPA Science Inventory

    As we enter the 21st century, policymakers face complex decisions regarding options for meeting the demand for transportation fuels. There is now a broad scientific consensus that the burning of fossil fuels has been contributing to climate change, and the transportation sector i...

  12. Outlook for alternative transportation fuels

    SciTech Connect

    Gushee, D.E.

    1996-12-31

    This presentation provides a brief review of regulatory issues and Federal programs regarding alternative fuel use in automobiles. A number of U.S. DOE initiatives and studies aimed at increasing alternative fuels are outlined, and tax incentives in effect at the state and Federal levels are discussed. Data on alternative fuel consumption and alternative fuel vehicle use are also presented. Despite mandates, tax incentives, and programs, it is concluded alternative fuels will have minimal market penetration. 7 refs., 5 tabs.

  13. Solid fuel applications to transportation engines

    SciTech Connect

    Rentz, Richard L.; Renner, Roy A.

    1980-06-01

    The utilization of solid fuels as alternatives to liquid fuels for future transportation engines is reviewed. Alternative liquid fuels will not be addressed nor will petroleum/solid fuel blends except for the case of diesel engines. With respect to diesel engines, coal/oil mixtures will be addressed because of the high interest in this specific application as a result of the large number of diesel engines currently in transportation use. Final assessments refer to solid fuels only for diesel engines. The technical assessments of solid fuels utilization for transportation engines is summarized: solid fuel combustion in transportation engines is in a non-developed state; highway transportation is not amenable to solid fuels utilization due to severe environmental, packaging, control, and disposal problems; diesel and open-cycle gas turbines do not appear worthy of further development, although coal/oil mixtures for slow speed diesels may offer some promise as a transition technology; closed-cycle gas turbines show some promise for solid fuels utilization for limited applications as does the Stirling engine for use of cleaner solid fuels; Rankine cycle engines show good potential for limited applications, such as for locomotives and ships; and any development program will require large resources and sophisticated equipment in order to advance the state-of-the-art.

  14. SEU43 fuel bundle shielding analysis during spent fuel transport

    SciTech Connect

    Margeanu, C. A.; Ilie, P.; Olteanu, G.

    2006-07-01

    The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)

  15. Alternatives to traditional transportation fuels 1993

    SciTech Connect

    Not Available

    1995-01-01

    In recent years, gasoline and diesel fuel have accounted for about 80 percent of total transportation fuel and nearly all of the fuel used in on-road vehicles. Growing concerns about the environmental effects of fossil fuel use and the Nation`s high level of dependence on foreign oil are providing impetus for the development of replacements or alternatives for these traditional transportation fuels. (The Energy Policy Act of 1992 definitions of {open_quotes}replacement{close_quotes} and {open_quotes}alternative{close_quotes} fuels are presented in the following box.) The Alternative Motor Fuels Act of 1988, the Clean Air Act Amendments of 1990 (CAAA90) and the Energy Policy Act of 1992 (EPACT) are significant legislative forces behind the growth of replacement fuel use. Alternatives to Traditional Transportation Fuels 1993 provides the number of on-road alternative fueled vehicles in use in the United States, alternative and replacement fuel consumption, and information on greenhouse gas emissions resulting from the production, delivery, and use of replacement fuels for 1992, 1993, and 1995.

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

    SciTech Connect

    Hertel, N.E.; Napolitano, D.G.

    1997-12-01

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

  17. Fuel cell system for transportation applications

    DOEpatents

    Kumar, R.; Ahmed, S.; Krumpelt, M.; Myles, K.M.

    1993-09-28

    A propulsion system is described for a vehicle having pairs of front and rear wheels and a fuel tank. An electrically driven motor having an output shaft operatively connected to at least one of said pair of wheels is connected to a fuel cell having a positive electrode and a negative electrode separated by an electrolyte for producing dc power to operate the motor. A partial oxidation reformer is connected both to the fuel tank and to the fuel cell and receives hydrogen-containing fuel from the fuel tank and uses water and air for partially oxidizing and reforming the fuel in the presence of an oxidizing catalyst and a reforming catalyst to produce a hydrogen-containing gas. The hydrogen-containing gas is sent from the partial oxidation reformer to the fuel cell negative electrode while air is transported to the fuel cell positive electrode to produce dc power for operating the electric motor. 3 figures.

  18. Fuel cell system for transportation applications

    DOEpatents

    Kumar, Romesh; Ahmed, Shabbir; Krumpelt, Michael; Myles, Kevin M.

    1993-01-01

    A propulsion system for a vehicle having pairs of front and rear wheels and a fuel tank. An electrically driven motor having an output shaft operatively connected to at least one of said pair of wheels is connected to a fuel cell having a positive electrode and a negative electrode separated by an electrolyte for producing dc power to operate the motor. A partial oxidation reformer is connected both to the fuel tank and to the fuel cell receives hydrogen-containing fuel from the fuel tank and water and air and for partially oxidizing and reforming the fuel with water and air in the presence of an oxidizing catalyst and a reforming catalyst to produce a hydrogen-containing gas. The hydrogen-containing gas is sent from the partial oxidation reformer to the fuel cell negative electrode while air is transported to the fuel cell positive electrode to produce dc power for operating the electric motor.

  19. Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez, J. S.

    2015-01-01

    [Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

  20. Implementation of ANSI/ASME NQA-1 for development of GA-4 and GA-9 spent fuel casks

    SciTech Connect

    Dunlap, M.G.

    1989-06-01

    The Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) has awarded General Atomics (GA) a contract to develop a legal weight truck (LWT) transportation system to transport boiling-water-reactor (BWR) and pressurized water reactor (PWR) spent fuels. DOE's contract requires GA to establish, maintain, and implement an NRC-approved Quality Assurance Program in accordance with ANSI/ASME NQA-1, ''Quality Assurance Program Requirements'' 1986 Edition (basic requirements and supplements), and the 1986 la Addenda. GA's Quality Assurance Program is a company-wide system described in GA's Quality Assurance Manual (QAM). General Atomics' Quality Assurance Program is implemented for a specific project by means of a Quality Assurance Program Document (QAPD). The QAPD is the document used by GA to invoke and document selected quality and project-applicable requirements. The QAPD is a tailored technical quality plan for a specific contract which can add to, revise, or delete certain portions of the basic GA QA Manual. The QAPD then becomes the controlling QA document for a specific project, and the provisions of the QAPD are followed by all project team members as though it were the QA Manual. For the legal weight truck (LWT) transportation project, GA has developed a QAPD that invokes the NRC-approved QA Program and also addresses the additional requirements that are included in the NQA-1 1986 Edition and la Addenda and the additional DOE/ID imposed requirements. 8 refs., 1 fig., 1 tab.

  1. Alternatives to traditional transportation fuels 1995

    SciTech Connect

    1996-12-01

    This report provides information on transportation fuels other than gasoline and diesel, and the vehicles that use these fuels. The Energy Information Administration (EIA) provides this information to support the U.S. Department of Energy`s reporting obligations under Section 503 of the Energy Policy Act of 1992 (EPACT). The principal information contained in this report includes historical and year-ahead estimates of the following: (1) the number and type of alterative-fueled vehicles (AFV`s) in use; (2) the consumption of alternative transportation fuels and {open_quotes}replacement fuels{close_quotes}; and (3) the number and type of alterative-fueled vehicles made available in the current and following years. In addition, the report contains some material on special topics. The appendices include a discussion of the methodology used to develop the estimates (Appendix A), a map defining geographic regions used, and a list of AFV suppliers.

  2. DOE perspective on fuel cells in transportation

    SciTech Connect

    Kost, R.

    1996-04-01

    Fuel cells are one of the most promising technologies for meeting the rapidly growing demand for transportation services while minimizing adverse energy and environmental impacts. This paper reviews the benefits of introducing fuel cells into the transportation sector; in addition to dramatically reduced vehicle emissions, fuel cells offer the flexibility than use petroleum-based or alternative fuels, have significantly greater energy efficiency than internal combustion engines, and greatly reduce noise levels during operation. The rationale leading to the emphasis on proton-exchange-membrane fuel cells for transportation applications is reviewed as are the development issues requiring resolution to achieve adequate performance, packaging, and cost for use in automobiles. Technical targets for power density, specific power, platinum loading on the electrodes, cost, and other factors that become increasingly more demanding over time have been established. Fuel choice issues and pathways to reduced costs and to a renewable energy future are explored. One such path initially introduces fuel cell vehicles using reformed gasoline while-on-board hydrogen storage technology is developed to the point of allowing adequate range (350 miles) and refueling convenience. This scenario also allows time for renewable hydrogen production technologies and the required supply infrastructure to develop. Finally, the DOE Fuel Cells in Transportation program is described. The program, whose goal is to establish the technology for fuel cell vehicles as rapidly as possible, is being implemented by means of the United States Fuel Cell Alliance, a Government-industry alliance that includes Detroit`s Big Three automakers, fuel cell and other component suppliers, the national laboratories, and universities.

  3. 30 CFR 75.1906 - Transport of diesel fuel.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Transport of diesel fuel. 75.1906 Section 75... diesel fuel. (a) Diesel fuel shall be transported only by diesel fuel transportation units or in safety... fuel storage facilities. (c) Safety cans that leak must be promptly removed from the mine. (d)...

  4. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    SciTech Connect

    Yuan, Y.C.; Chen, S.Y.; LePoire, D.J.; Rothman, R.

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  5. 30 CFR 75.1906 - Transport of diesel fuel.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Transport of diesel fuel. 75.1906 Section 75... HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Diesel-Powered Equipment § 75.1906 Transport of diesel fuel. (a) Diesel fuel shall be transported only by diesel fuel transportation units or in...

  6. 30 CFR 75.1906 - Transport of diesel fuel.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Transport of diesel fuel. 75.1906 Section 75... HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Diesel-Powered Equipment § 75.1906 Transport of diesel fuel. (a) Diesel fuel shall be transported only by diesel fuel transportation units or in...

  7. 30 CFR 75.1906 - Transport of diesel fuel.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Transport of diesel fuel. 75.1906 Section 75... HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Diesel-Powered Equipment § 75.1906 Transport of diesel fuel. (a) Diesel fuel shall be transported only by diesel fuel transportation units or in...

  8. Analysis of the risk of transporting spent nuclear fuel by train

    SciTech Connect

    Elder, H.K.

    1981-09-01

    This report uses risk analyses to analyze the safety of transporting spent nuclear fuel for commercial rail shipping systems. The rail systems analyzed are those expected to be used in the United States when the total electricity-generating capacity by nuclear reactors is 100 GW in the late 1980s. Risk as used in this report is the product of the probability of a release of material to the environment and the consequences resulting from the release. The analysis includes risks in terms of expected fatalities from release of radioactive materials due to transportation accidents involving PWR spent fuel shipped in rail casks. The expected total risk from such shipments is 1.3 x 10/sup -4/ fatalities per year. Risk spectrums are developed for shipments of spent fuel that are 180 days and 4 years out-of-reactor. The risk from transporting spent fuel by train is much less (by 2 to 4 orders of magnitude) than the risk to society from other man-caused events such as dam failure.

  9. Fuel cell assembly with electrolyte transport

    DOEpatents

    Chi, Chang V.

    1983-01-01

    A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

  10. Spent Nuclear Fuel Transport Reliability Study

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    2016-01-01

    This conference paper was orignated and shorten from the following publisehd PTS documents: 1. Jy-An Wang, Hao Jiang, and Hong Wang, Dynamic Deformation Simulation of Spent Nuclear Fuel Assembly and CIRFT Deformation Sensor Stability Investigation, ORNL/SPR-2015/662, November 2015. 2. Jy-An Wang, Hong Wang, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, ORNL/TM-2014/214, May 2015. 3. Jy-An Wang, Hong Wang, Hao Jiang, Yong Yan, Bruce Bevard, Spent Nuclear Fuel Vibration Integrity Study 16332, WM2016 Conference, March 6 10, 2016, Phoenix, Arizona.

  11. Status update of the BWR cask simulator

    SciTech Connect

    Lindgren, Eric R.; Durbin, Samuel G.

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of

  12. Challenges for fuel cells in transport applications

    NASA Astrophysics Data System (ADS)

    Chalk, Steven G.; Miller, James F.; Wagner, Fred W.

    The US Department of Energy (DOE) and the US automotive industry are working cooperatively under the auspices of the Partnership for a New Generation of Vehicles (PNGV) to develop a six-passenger automobile that can achieve up to 80 miles/gal. These partners are continuing to invest heavily in research and development of polymer electrolyte membrane (PEM) fuel cells as a clean and efficient alternative technology for transport applications. In the past few years, US automakers have made significant advances in fuel cell technology and have announced plans to put fuel cell vehicles on the market by 2004. DOE is working with industry suppliers to address some of the biggest remaining challenges, which include fuel processing and lowering the cost of fuel cell systems. The refueling infrastructure necessary to support fuel cell vehicles presents additional unresolved issues. This paper provides a status report on the PNGV program and provides an overview of the technical accomplishments and future plans of the DOE Fuel Cells for Transportation Program.

  13. 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.

  14. Convert natural gas into clean transportation fuels

    SciTech Connect

    Agee, M.A.

    1997-03-01

    A new process economically converts natural gas into synthetic transportation fuels that are free of sulfur, metals, aromatics and are clear in appearance. The process, developed by Syntroleum Corp., is energy self-sufficient and can be implemented in sizes small enough to fit a large number of the world`s gas fields. The process is described.

  15. Fuel cell power plants for transportation applications

    SciTech Connect

    Huff, J.R.

    1991-12-31

    Over the past 35 years, the transportation sector has accounted fr approximately 25% of the total gross energy consumption in the United States. As the largest energy user in the United States, transportation accounts for approximately 66% of the country`s current petroleum consumption. Fuel cell power plants using nonpetroleum fuels such as methanol could significantly reduce US dependency on petroleum resources. They offer the additional advantage of minimal air pollution thereby addressing another issue of major concern in the US fuel cell power plant use in city buses and other vehicles is being explored in a number of US Department of Energy and industrial programs that will be described in this paper. 5 refs.

  16. Shipping Cask Design Review Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1998-01-04

    Version 01 SCANS (Shipping Cask ANalysis System) is a microcomputer based system of computer programs and databases for evaluating safety analysis reports on spent fuel shipping casks. SCANS calculates the global response to impact loads, pressure loads, and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. Analysis options are based on regulatory cases described in the Code of Federal Regulations (1983) and Regulatory Guides published by the NRC in 1977more » and 1978. The system is composed of a series of menus and input entry cask analysis, and output display programs. An analysis is performed by preparing the necessary input data and then selecting the appropriate analysis: impact, thermal (heat transfer), thermally-induced stress, or pressure-induced stress. All data are entered through input screens with descriptive data requests, and, where possible, default values are provided. Output (i.e., impact force, moment and sheer time histories; impact animation; thermal/stress geometry and thermal/stress element outlines; temperature distributions as isocontours or profiles; and temperature time histories) is displayed graphically and can also be printed.« less

  17. 76 FR 12825 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Confirmation of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-09

    ... January 13, 2011 (76 FR 2243). This direct final rule amended the NRC's spent fuel storage regulations at...: Gregory.Trussell@nrc.gov . SUPPLEMENTARY INFORMATION: On January 13, 2011 (76 FR 2243), the NRC published... reconstituted fuel assemblies; add requirements to qualify metal matrix composite neutron absorbers...

  18. Spent nuclear fuel project product specification

    SciTech Connect

    Pajunen, A.L.

    1998-01-30

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification.

  19. 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.

  20. Technical Review Report for the Justification for Shipment of Sodium-Bonded Carbide Fuel Pins in the T-3 Cask

    SciTech Connect

    West, M; DiSabatino, A

    2008-01-04

    This report documents the review of the Fluor Submittal (hereafter, the Submittal), prepared by Savannah River Packaging Technology (SRPT) of Savannah River National Laboratory (SRNL), at the request of the Department of Energy's (DOE) Richland Operations Office, for the shipment of unirradiated and irradiated sodium-bonded carbide fuel pins. The sodium-bonded carbide fuel pins are currently stored at the Fast Flux Test Facility (FFTF) awaiting shipment to Idaho National Laboratory (INL). Normally, modified contents are included into the next revision of the SARP. However, the contents, identified to be shipped from FFTF to Idaho National Laboratory, are a one-way shipment of 18 irradiated fuel pins and 7 unirradiated fuel pins, where the irradiated and unirradiated fuel pins are shipped separately, and can be authorized with a letter amendment to the existing Certificate of Compliance (CoC).

  1. Topical safety analysis report for the transportation of the NUHOMS{reg_sign} dry shielded canister. Volume 1

    SciTech Connect

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS{reg_sign}) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS{reg_sign} DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS{reg_sign} Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport.

  2. Fuel cell development for transportation: Catalyst development

    SciTech Connect

    Doddapaneni, N.; Ingersoll, D.

    1996-12-31

    Fuel cells are being considered as alternative power sources for transportation and stationary applications. The degradation of commonly used electrode catalysts (e.g. Pt, Ag, and others) and corrosion of carbon substrates are making commercialization of fuel cells incorporating present day technologies economically problematic. Furthermore, due to the instability of the Pt catalyst, the performance of fuel cells declines on long-term operation. When methanol is used as the fuel, a voltage drop, as well as significant thermal management problems can be encountered, the later being due to chemical oxidation of methanol at the platinized carbon at the cathode. Though extensive work was conducted on platinized electrodes for both the oxidation and reduction reactions, due to the problems mentioned above, fuel cells have not been fully developed for widespread commercial use. Several investigators have previously evaluated metal macrocyclic complexes as alternative catalysts to Pt and Pt/Ru in fuel cells. Unfortunately, though they have demonstrated catalytic activity, these materials were found to be unstable on long term use in the fuel cell environment. In order to improve the long-term stability of metal macrocyclic complexes, we have chemically bonded these complexes to the carbon substrate, thereby enhancing their catalytic activity as well as their chemical stability in the fuel cell environment. We have designed, synthesized, and evaluated these catalysts for O{sub 2} reduction, H{sub 2} oxidation, and direct methanol oxidation in Proton Exchange Membrane (PEM) and aqueous carbonate fuel cells. These catalysts exhibited good catalytic activity and long-term stability. In this paper we confine our discussion to the initial performance results of some of these catalysts in H{sub 2}/O{sub 2} PEM fuel cells, including their long-term performance characteristics as well as CO poisoning effects on these catalysts.

  3. The transportation operations system: A description

    SciTech Connect

    Best, R.E.; Danese, F.L.; Dixon, L.D.; Peterson, R.W. ); Pope, R.B. )

    1990-01-01

    This paper presents a description of the system for transporting radioactive waste that may be deployed to accomplish the assigned system mission, which includes accepting spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from waste generator sites and transporting them to the FWMS destination facilities. The system description presented here contains, in part, irradiated fuel and waste casks, ancillary equipments, truck, rail, and barge transporters, cask and vehicle traffic management organizations, maintenance facilities, and other operations elements. The description is for a fully implemented system, which is not expected to be achieved, however, until several years after initial operations. 6 figs.

  4. Hydrogen fuel dispensing station for transportation vehicles

    SciTech Connect

    Singh, S.P.N.; Richmond, A.A.

    1995-07-01

    A technical and economic assessment is being conducted of a hydrogen fuel dispensing station to develop an understanding of the infrastructure requirements for supplying hydrogen fuel for mobile applications. The study includes a process design of a conceptual small-scale, stand-alone, grassroots fuel dispensing facility (similar to the present-day gasoline stations) producing hydrogen by steam reforming of natural gas. Other hydrogen production processes (such as partial oxidation of hydrocarbons and water electrolysis) were reviewed to determine their suitability for manufacturing the hydrogen. The study includes an assessment of the environmental and other regulatory permitting requirements likely to be imposed on a hydrogen fuel dispensing station for transportation vehicles. The assessment concludes that a dispensing station designed to produce 0.75 million standard cubic feet of fuel grade (99.99%+ purity) hydrogen will meet the fuel needs of 300 light-duty vehicles per day. Preliminary economics place the total capital investment (in 1994 US dollars) for the dispensing station at $4.5 million and the annual operating costs at around $1 million. A discounted cash-flow analysis indicates that the fuel hydrogen product price (excluding taxes) to range between $1.37 to $2.31 per pound of hydrogen, depending upon the natural gas price, the plant financing scenario, and the rate of return on equity capital. A report on the assessment is due in June 1995. This paper presents a summary of the current status of the assessment.

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

    SciTech Connect

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

    2013-04-01

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

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

    SciTech Connect

    Bohachek, Randy; Wallace, Bruce; Winston, Phil; Marschman, Steve

    2013-04-30

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

  7. Multi-fuel reformers for fuel cells used in transportation. Phase 1: Multi-fuel reformers

    NASA Astrophysics Data System (ADS)

    1994-05-01

    DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

  8. A Cask Processing Enclosure for the TRU Waste Processing Center - 13408

    SciTech Connect

    Newman, John T.; Mendez, Nicholas

    2013-07-01

    This paper will discuss the key elements considered in the design, construction, and use of an enclosure system built for the TRU Waste Processing Center (TWPC). The TWPC system is used for the repackaging and volume reduction of items contaminated with radioactive material, hazardous waste and mixed waste. The modular structural steel frame and stainless steel skin was designed for rapid field erection by the use of interchangeable self-framing panel sections to allow assembly of a sectioned containment building and for ease of field mobility. The structure was installed on a concrete floor inside of an outer containment building. The major sections included an Outer Cask Airlock, Inner Cask Airlock, Cask Process Area, and Personnel Airlocks. Casks in overpacks containing transuranic waste are brought in via an inter-site transporter. The overpack lid is removed and the cask/overpack is transferred into the Outer Cask Airlock. A contamination cover is installed on the overpack body and the Outer Cask Airlock is closed. The cask/overpack is transferred into the Inner Cask Airlock on a cask bogie and the Inner Cask Airlock is closed. The cask lid is removed and the cask is transferred into the Cask Process Area where it is placed on a cask tilting station. Once the Cask Processing Area is closed, the cask tilt station is activated and wastes are removed, size reduced, then sorted and re-packaged into drums and standard waste boxes through bag ports. The modular system was designed and built as a 'Fast Track' project at IP Systems in Broomfield Colorado and then installed and is currently in use at the DOE TWPC located near Oak Ridge, Tennessee. (authors)

  9. Ultrasonic cleaning of fuel assemblies

    SciTech Connect

    Kondoh, Keisuke; Fujita, Chitoshi; Sakai, Hitoshi

    1994-12-31

    During fuel transportation, contamination of the transfer cask can lead to radiation dosage. That is radioactive crud becomes detached from the fuel surface and is deposited inside the cask. To avoid this at the Tsuruga Power Station Unit 1, crud was removed from fuel assemblies in advance of fuel transportation work. An ultrasonic cleaning process was adopted for this purpose; ultrasonic methods excel over other methods for this type of cleaning. Our process is also able to clean fuel assemblies without removing the channel box. Since this is the first time that the ultrasonic method was applied to fuel assemblies at the light water reactor in Japan on a large scale, the efficiency and the impact on plant instrumentation of the method were examined by performing preliminary test. Based on these tests, an optimum cleaning procedure was established.

  10. 76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-28

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181..., 1997, and published in the Federal Register on September 3, 1997 (62 FR 46517), this rule is classified... June 10, 1998 (63 FR 31883), directed that the Government's documents be in clear and...

  11. Nuclear Energy and Synthetic Liquid Transportation Fuels

    NASA Astrophysics Data System (ADS)

    McDonald, Richard

    2012-10-01

    This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

  12. Safety evaluation for packaging (onsite) SERF cask

    SciTech Connect

    Edwards, W.S.

    1997-10-24

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

  13. FFTF disposable solid waste cask

    SciTech Connect

    Thomson, J. D.; Goetsch, S. D.

    1983-01-01

    Disposal of radioactive waste from the Fast Flux Test Facility (FFTF) will utilize a Disposable Solid Waste Cask (DSWC) for the transport and burial of irradiated stainless steel and inconel materials. Retrievability coupled with the desire for minimal facilities and labor costs at the disposal site identified the need for the DSWC. Design requirements for this system were patterned after Type B packages as outlined in 10 CFR 71 with a few exceptions based on site and payload requirements. A summary of the design basis, supporting analytical methods and fabrication practices developed to deploy the DSWC is provided in this paper.

  14. Final Hazard Classification for the FFTF Solid Waste Cask

    SciTech Connect

    HIMES, D.A.

    2002-07-03

    The Solid Waste Cask (SWC) (a major component of the Fast Flux Test Facility (FFTF) spent fuel offload system) is a shielded, bottom-loading cask containing an internal hoist system used to transfer irradiated fuel or non-fuel components from the Interim Examination and Maintenance Cell (IEM Cell) to the Cask Loading Station (CLS). The SWC is assumed to be loaded with 7 irradiated fuel assemblies in a Core Component Container (CCC) having maximum average burn-ups of 150,000 MWd/MTHM. Results show that the fuel handling activities with the SWC loaded with 7 irradiated fuel assemblies in a CCC should be classified as a Category 3 hazard. This conclusion is consistent with the relative simplicity of the system and passive nature of the barriers for purposes of determining the graded approach specified in DOE-STD-1027-92 (DOE 1992).

  15. ACCIDENTAL DROP OF A CARBON STEEL/LEAD SHIPPING CASK AT LOW TEMPERATURES

    SciTech Connect

    B. D. Hawkes; K. R. Durstine

    2007-07-01

    A shielded cask is used to transport radioactive materials between facilities. The cask was fabricated with an outer and inner shell of hot rolled low carbon steel. Lead was poured in the annular space between the shells to provide radiation shielding. Carbon steel is known to be susceptible to lowtemperature brittle fracture under impact loading. This paper will present the analysis results representing postulated transportation accidents during on-site transfers of the cask. The accident scenarios were based on a series of cask drops onto a rigid surface from a height of 6 ft assuming brittle failure of the cask shell at subzero temperatures. Finite element models of the cask and its contents were solved and post processed using ABAQUS software. Each model was examined for failure to contain radioactive materials and/or significant loss of radiation shielding. Results of these analyses show that the body of the cask exhibits considerable ruggedness and will remain largely intact after the impact. There will be deformation of the main cask body with localized brittle failure of the cask outer shell and components and but no complete penetration of the cask shielding. The cask payload outer waste can will experience some permanent plastic deformation in each drop, but will not be deformed to the point where it will rupture, thus maintaining confinement of the can contents.

  16. Consolidation and disposal of PWR fuel inserts

    SciTech Connect

    Wakeman, B.H. )

    1992-08-01

    Design and licensing of the Surry Power Station Independent Spent Fuel Storage Installation was initiated in 1982 by Virginia Power as part of a comprehensive strategy to increase spent fuel storage capacity at the Station. Designed to use large, metal dry storage casks, the Surry Installation will accommodate 84 such casks with a total storage capacity of 811 MTU of spent pressurized water reactor fuel assemblies. Virginia Power provided three storage casks for testing at the Idaho National Engineerinq Laboratory's Test Area North and the testing results have been published by the Electric Power Research Institute. Sixty-nine spent fuel assemblies were transported in truck casks from the Surry Power Station to Test Area North for testing in the three casks. Because of restrictions imposed by the cask testing equipment at Test Area North, the irradiated insert components stored in these fuel assemblies at Surry were removed prior to transport of the fuel assemblies. Retaining these insert components proved to be a problem because of a shortage of spent fuel assemblies in the spent fuel storage pool that did not already contain insert components. In 1987 Virginia Power contracted with Chem-Nuclear Systems, Inc. to process and dispose of 136 irradiated insert components consisting of 125 burnable poison rod assemblies, 10 thimble plugging devices and 1 part-length rod cluster control assembly. This work was completed in August and September 1987, culminating in the disposal at the Barnwell, SC low-level radioactive waste facility of two CNS 3-55 liners containing the consolidated insert components.

  17. Nevada commercial spent nuclear fuel transportation experience

    SciTech Connect

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed.

  18. Sustainable fuel for the transportation sector

    SciTech Connect

    Agrawal, R.; Singh, N.R.; Ribeiro, F.H.; Delgass, W.N.

    2007-03-20

    A hybrid hydrogen-carbon (H{sub 2}CAR) process for the production of liquid hydrocarbon fuels is proposed wherein biomass is the carbon source and hydrogen is supplied from carbon-free energy. To implement this concept, a process has been designed to co-feed a biomass gasifier with H{sub 2} and CO{sub 2} recycled from the H{sub 2}-CO to liquid conversion reactor. Modeling of this biomass to liquids process has identified several major advantages of the H{sub 2}CAR process. The land area needed to grow the biomass is <40% of that needed by other routes that solely use biomass to support the entire transportation sector. Whereras the literature estimates known processes to be able to produce {approx}30% of the United States transportation fuel from the annual biomass of 1.366 billion tons, the H{sub 2}CAR process shows the potential to supply the entire United States transportation sector from that quantity of biomass. The synthesized liquid provides H{sub 2} storage in an open loop system. Reduction to practice of the H{sub 2}CAR route has the potential to provide the transportation sector for the foreseeable future, using the existing infrastructure. The rationale of using H{sub 2} in the H{sub 2}CAR process is explained by the significantly higher annualized average solar energy conversion efficiency for hydrogen generation versus that for biomass growth. For coal to liquids, the advantage of H{sub 2}CAR is that there is no additional CO{sub 2} release to the atmosphere due to the replacement of petroleum with coal, thus eliminating the need to sequester CO{sub 2}.

  19. Sustainable fuel for the transportation sector

    PubMed Central

    Agrawal, Rakesh; Singh, Navneet R.; Ribeiro, Fabio H.; Delgass, W. Nicholas

    2007-01-01

    A hybrid hydrogen-carbon (H2CAR) process for the production of liquid hydrocarbon fuels is proposed wherein biomass is the carbon source and hydrogen is supplied from carbon-free energy. To implement this concept, a process has been designed to co-feed a biomass gasifier with H2 and CO2 recycled from the H2-CO to liquid conversion reactor. Modeling of this biomass to liquids process has identified several major advantages of the H2CAR process. (i) The land area needed to grow the biomass is <40% of that needed by other routes that solely use biomass to support the entire transportation sector. (ii) Whereas the literature estimates known processes to be able to produce ≈30% of the United States transportation fuel from the annual biomass of 1.366 billion tons, the H2CAR process shows the potential to supply the entire United States transportation sector from that quantity of biomass. (iii) The synthesized liquid provides H2 storage in an open loop system. (iv) Reduction to practice of the H2CAR route has the potential to provide the transportation sector for the foreseeable future, using the existing infrastructure. The rationale of using H2 in the H2CAR process is explained by the significantly higher annualized average solar energy conversion efficiency for hydrogen generation versus that for biomass growth. For coal to liquids, the advantage of H2CAR is that there is no additional CO2 release to the atmosphere due to the replacement of petroleum with coal, thus eliminating the need to sequester CO2. PMID:17360377

  20. Modifications to SAS4 to provide cask dose rate profiles

    SciTech Connect

    Napolitano, D.G.; Sweezy, J.E.; Henkel, C.S.

    1997-12-01

    SAS4 of the SCALE code system has been used extensively by NAC International (NAC) to perform storage and transport cask shielding analyses. SAS4 utilizes a one-dimensional XSDRNPM adjoint calculation of the cask to generate biasing parameters for a three-dimensional MORSE-SGC Monte Carlo model of the cask geometry. This technique is very efficient in getting particles to tally at the cask exterior surfaces. However, SAS4/MORSE-SGC is limited to the use of point detectors (next-event estimators) and large surface detectors (surface-crossing estimators). Modifications to SAS4 were made to allow a more flexible use of the surface detectors. This modification allows multiple nonoverlapping surface detectors on each surface and allows each surface detector to be broken into subdetectors. The use of subdetectors enables the user to obtain detailed surface dose rate profiles. Tallies can now be performed on all surfaces of the cask and at user-specified distances from the cask surface. The subdetectors provide an alternative to point detectors and excessive computational time. The NAC version of SAS4 is called SAS4A. A comparison of CPU time and dose rates is made between SAS4 point detectors and SAS4A surface subdetection on the NLI {1/2} transport cask.

  1. Casks (computer analysis of storage casks): A microcomputer based analysis system for storage cask review

    SciTech Connect

    Chen, T.F.; Mok, G.C.; Carlson, R.W.

    1995-08-01

    CASKS is a microcomputer based computer system developed by LLNL to assist the Nuclear Regulatory Commission in performing confirmatory analyses for licensing review of radioactive-material storage cask designs. The analysis programs of the CASKS computer system consist of four modules: the impact analysis module, the thermal analysis module, the thermally-induced stress analysis module, and the pressure-induced stress analysis module. CASKS uses a series of menus to coordinate input programs, cask analysis programs, output programs, data archive programs and databases, so the user is able to run the system in an interactive environment. This paper outlines the theoretical background on the impact analysis module and the yielding surface formulation. The close agreement between the CASKS analytical predictions and the results obtained form the two storage casks drop tests performed by SNL and by BNFL at Winfrith serves as the validation of the CASKS impact analysis module.

  2. Thermal evaluation facility for LMFBR spent fuel transport

    SciTech Connect

    Wesley, D.A.

    1980-04-01

    A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations.

  3. Shielding and Containment Evaluations of the NAC-LWT Cask with Tritium Burnable Poison Rods

    SciTech Connect

    Holger Pfeifer; Norman Meinert

    2000-06-04

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the U.S. Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for a shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload.

  4. Shielding and containment evaluations of the NAC-LWT cask with tritium burnable poison rods

    SciTech Connect

    Pfeifer, H.; Meinert, N.

    2000-07-01

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the US Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for the shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload.

  5. Evaluation of gamma radiation shielding for nuclear waste shipping casks

    SciTech Connect

    Liu, Y.Y.; Carlson, R.D.; Primeau, S.J.; Wangler, M.E.

    1998-05-01

    A method has been developed for evaluating gamma radiation shielding of shipping casks that are used to transport nuclear waste with ill-defined radionuclide contents. The method is based on calculations that establish individual limits for a comprehensive list of radionuclides in the waste, assuming that each radionuclide is uniformly distributed in a volumetric source in the cask. For multiple radionuclide mixtures, a linear fraction rule is used to restrict the total amount of radionuclides such that the sum of the fractions does not exceed 1. As long as the radionuclide limits and the linear fraction rule are followed, it can be shown that the regulatory dose rate requirements for a cask will be satisfied under normal conditions of transport and in a hypothetical accident during which the shielding thickness of the cask has been reduced by 40%.

  6. 49 CFR 175.310 - Transportation of flammable liquid fuel; aircraft only means of transportation.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 2 2013-10-01 2013-10-01 false Transportation of flammable liquid fuel; aircraft only means of transportation. 175.310 Section 175.310 Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION HAZARDOUS MATERIALS REGULATIONS CARRIAGE...

  7. 49 CFR 175.310 - Transportation of flammable liquid fuel; aircraft only means of transportation.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Transportation of flammable liquid fuel; aircraft only means of transportation. 175.310 Section 175.310 Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION HAZARDOUS MATERIALS REGULATIONS CARRIAGE...

  8. High fuel price: Will Indonesian shift to public transportation?

    NASA Astrophysics Data System (ADS)

    Sopha, Bertha Maya; Pamungkas, Adhiguna Ramadhani

    2016-06-01

    Public transportation has been declining over years, while on the other hand, private vehicles are dramatically increasing. The share of public transportation was 38.3% in 2002 and slowly decreasing to 12.9% in 2010. Cheap fuel price has been alleged to be the main cause for the increased private vehicles. The declining trend of public transportation needs further investigation whether higher fuel price indeed influences the choice of transportation mode. The present study therefore aims at exploring the preference of using public transportation compared to motorcycle and private car for various fuel price and identifying barriers toward public transportation. A survey was conducted in 2013 to capture the preference of each transportation mode given different fuel price. A questionnaire which was designed according to the structure of Analytical Hierarchy Process (AHP) was distributed using random sampling in ten cities in Sumatra and Java islands, Indonesia. Results indicate that the increased fuel price would not lead to significant increase of public transportation users. Motorcycle seems continuously being the dominating transportation mode in the future. On the other hand, issues resulted from limited public transportation capacity such as long travel time, security and safety issues, limited route, poor schedule appear to be the most barriers of using public transportation. It is implied that in order to promote public transportation, interventions should be introduced simultaneously at both supply (i.e., increasing public transportation capacity) and demand (i.e., high fuel price) sides. Limitations of the study are also discussed.

  9. THERMAL MODELING ANALYSIS OF SRS 70 TON CASK

    SciTech Connect

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

    2011-03-08

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

  10. Spreader beam analysis for the CASTOR GSF cask

    SciTech Connect

    Clements, E.P.

    1997-04-07

    The purpose of this report is to document the results of the 150% rated capacity load test performed by DynCorp Hoisting and Rigging on the CASTOR GSF special cask lifting beams. The two lifting beams were originally rated and tested at 20,000kg (44,000lb) by the cask manufacturer in Germany. The testing performed by DynCorp rated and tested the lifting beams to 30,000 kg (66,000 lb) +0%, -5%, for Hanford Site use. The CASTOR GSF cask, used to transport isotopic Heat Sources (canisters), must be lifted with its own designed lifting beam system (Figures 1, 2, and 3). As designed, the beam material is RSt 37-2 (equivalent to American Society for Testing and Materials [ASTM] A-570), the eye plate is St 52-2 (equivalent to ASTM A-516), and the lifting pin is St 50 (equivalent to ASTM A-515). The beam has two opposing 58 mm (2.3 in.) diameter by 120 mm(4.7 in.) length, high grade steel pins that engage the cask for lifting. The pins have a manual locking mechanism to prevent disengagement from the casks. The static, gross weight (loaded) of the cask 18,640 kg (41,000 lb) on the pins prevents movement of the pins during lifting. This is due to the frictional force of the cask on the pins when lifting begins.

  11. CASKS (Computer Analysis of Storage casKS): A microcomputer based analysis system for storage cask design review. User`s manual to Version 1b (including program reference)

    SciTech Connect

    Chen, T.F.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L.; Mok, G.C.

    1995-02-01

    CASKS (Computer Analysis of Storage casKS) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent-fuel storage casks. The bulk of the complete program and this user`s manual are based upon the SCANS (Shipping Cask ANalysis System) program previously developed at LLNL. A number of enhancements and improvements were added to the original SCANS program to meet requirements unique to storage casks. CASKS is an easy-to-use system that calculates global response of storage casks to impact loads, pressure loads and thermal conditions. This provides reviewers with a tool for an independent check on analyses submitted by licensees. CASKS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests.

  12. Accidental Drop of a Carbon Steel/Lead Shipping Cask (HFEF 14) at Low Temperatures

    SciTech Connect

    Brian D. Hawkes; Michael E. Nitzel

    2007-08-01

    A shielded cask is used to transport radioactive materials between facilities at the Idaho National Laboratory. The cask was fabricated with an outer and inner shell of A36 carbon steel with lead poured in the annular space between the shells to provide radiation shielding. Carbon steel is known to be susceptible to low-temperature brittle fracture under impact loading. This paper will present the analysis results representing postulated transportation accidents during on-site transfers of the cask at subzero temperatures. The accident scenarios were based on a series of cask drops onto a rigid surface from a height of 1.83m (6 ft.) Finite element models of the cask and its contents were solved and post processed using the ABAQUS software. Each model was examined for failure to contain radioactive materials and/or significant loss of radiation shielding. Results of these analyses show that the body of the cask exhibits considerable ruggedness and will remain largely intact after the impact. There will be deformation of the main cask body with localized brittle failure of the cask outer shell and door structure. The cask payload outer waste can remains in the cask but will experience some permanent plastic deformation in each drop. It will not be deformed to the point where it will rupture, thus maintaining confinement of the can contents.

  13. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    SciTech Connect

    Wagner, J. C.

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  14. EPAct Alternative Fuel Transportation Program: Success Story (Fact Sheet)

    SciTech Connect

    Not Available

    2010-08-01

    This success story highlights the EPAct Alternative Fuel Transportation Program's series of workshops that bring fleets regulated under the Energy Policy Act of 1992 (EPAct) together with Clean Cities stakeholders and fuel providers to form and strengthen regional partnerships and initiate projects that will deploy more alternative fuel infrastructure.

  15. PADD 1 and PADD 3 Transportation Fuels Markets

    EIA Publications

    2016-01-01

    This study examines supply, consumption, and distribution of transportation fuels in Petroleum Administration for Defense Districts (PADDs) 1 and 3, or the U.S. East Coast and the Gulf Coast, respectively. The East Coast region includes states from Maine to Florida along the U.S. Atlantic Coast. The Gulf Coast region comprises states between New Mexico in the west to Alabama in the east along the Gulf of Mexico. For this study, transportation fuels include gasoline, diesel fuel and jet fuel. Residual fuel oil supply is also analyzed where applicable.

  16. Verification of maximum impact force for interim storage cask for the Fast Flux Testing Facility

    SciTech Connect

    Chen, W.W.; Chang, S.J.

    1996-06-01

    The objective of this paper is to perform an impact analysis of the Interim Storage Cask (ISC) of the Fast Flux Test Facility (FFTF) for a 4-ft end drop. The ISC is a concrete cask used to store spent nuclear fuels. The analysis is to justify the impact force calculated by General Atomics (General Atomics, 1994) using the ILMOD computer code. ILMOD determines the maximum force developed by the concrete crushing which occurs when the drop energy has been absorbed. The maximum force, multiplied by the dynamic load factor (DLF), was used to determine the maximum g-level on the cask during a 4-ft end drop accident onto the heavily reinforced FFTF Reactor Service Building`s concrete surface. For the analysis, this surface was assumed to be unyielding and the cask absorbed all the drop energy. This conservative assumption simplified the modeling used to qualify the cask`s structural integrity for this accident condition.

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

    SciTech Connect

    1997-01-01

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

  18. The NINO (No Inspector, No Operator system) cask-loading safeguards system

    SciTech Connect

    Fiarman, S.

    1987-01-01

    It is, in general difficult to determine by means of camera-surveillance techniques what is loaded into spent-fuel casks being prepared for shipment from light-water reactors to other reactors, reprocessing facilities, or long-term storage. Furthermore, the expected high frequency of cask loadings in the coming years would place too great a burden on the IAEA and Euratom inspectorates if each had to be observed by an inspector. For the case of shipment to other reactors and reprocessing facilities, the casks are soon opened and, in principle, their contents could be ascertained by direct inspection. In the case of long-term-storage facilities, the casks would stay sealed for years, thereby requiring the IAEA to know positively how many spent-fuel assemblies were loaded at the reactor and to have a continuity of knowledge of the cask's contents. It has been proposed instead that the facility operator place the cask seal on the cask within the field of view of a surveillance system linked to the cask seal. This solution, however, may not provide enough credibility for acceptance by the safeguards community. This paper presents an alternative to both inspector presence at cask loading and operator assistance in applying seals; this alternative is called the No Inspector, No Operator system (NINO).

  19. Potential of hydrogen fuel for future air transportation systems.

    NASA Technical Reports Server (NTRS)

    Small, W. J.; Fetterman, D. E.; Bonner, T. F., Jr.

    1973-01-01

    Recent studies have shown that hydrogen fuel can yield spectacular improvements in aircraft performance in addition to its more widely discussed environmental advantages. The characteristics of subsonic, supersonic, and hypersonic transport aircraft using hydrogen fuel are discussed, and their performance and environmental impact are compared to that of similar aircraft using conventional fuel. The possibilities of developing hydrogen-fueled supersonic and hypersonic vehicles with sonic boom levels acceptable for overland flight are also explored.

  20. Recent experience in planning, packaging and preparing non-commercial spent fuel for shipment within the United States

    SciTech Connect

    Johnson, P.E.; Shappert, L.B.; Turner, D.W.

    1996-06-01

    US DOE orders dictate that the aluminium clad fuels now stored at ORNL will be shipped to the Savannah River Site. A number of activities had to be carried out in order to ready the fuel for shipping, including choosing a cask capable of transporting the fuel, repackaging the fuel, developing a transportation plan, identifying the appropriate routes, and carrying out a readiness self assessment. These tasks have been successfully completed and are discussed herein.

  1. Chemical Kinetic Modeling of Advanced Transportation Fuels

    SciTech Connect

    PItz, W J; Westbrook, C K; Herbinet, O

    2009-01-20

    Development of detailed chemical kinetic models for advanced petroleum-based and nonpetroleum based fuels is a difficult challenge because of the hundreds to thousands of different components in these fuels and because some of these fuels contain components that have not been considered in the past. It is important to develop detailed chemical kinetic models for these fuels since the models can be put into engine simulation codes used for optimizing engine design for maximum efficiency and minimal pollutant emissions. For example, these chemistry-enabled engine codes can be used to optimize combustion chamber shape and fuel injection timing. They also allow insight into how the composition of advanced petroleum-based and non-petroleum based fuels affect engine performance characteristics. Additionally, chemical kinetic models can be used separately to interpret important in-cylinder experimental data and gain insight into advanced engine combustion processes such as HCCI and lean burn engines. The objectives are: (1) Develop detailed chemical kinetic reaction models for components of advanced petroleum-based and non-petroleum based fuels. These fuels models include components from vegetable-oil-derived biodiesel, oil-sand derived fuel, alcohol fuels and other advanced bio-based and alternative fuels. (2) Develop detailed chemical kinetic reaction models for mixtures of non-petroleum and petroleum-based components to represent real fuels and lead to efficient reduced combustion models needed for engine modeling codes. (3) Characterize the role of fuel composition on efficiency and pollutant emissions from practical automotive engines.

  2. Liquid transportation fuels from algal oils

    NASA Astrophysics Data System (ADS)

    Chen, Daichuan

    Liquid transportation fuels from renewable sources are becoming more prominent and important in modem society. Processing of hydrocarbon oils from algae has not been studied in detail in the past, so components which have been proposed for incorporation in algal oils via genetic engineering, such as cuparene, farnesene, phytol and squalene, have been subjected to processing via catalytic cracking in a pulse reactor at different temperatures. The cracking results showed that liquid products contained numerous high octane molecules which make it feasible for use in automobiles. Additionally, canola oil, chosen as an algal oil model compound, was studied as a feed for catalytic cracking in a fixed-bed reactor at atmospheric pressure over different types of zeolites. The results showed that MFI catalysts gave the highest yield of gasoline range products and lowest coke formation. Gallium loaded MFI zeolites increased the total aromatics yield for the canola oil cracking relative to the acid form of the zeolite. Finally, algal oils were cracked on several selected zeolites, and the results showed the same trend as canola oil cracking. MFI gave the highest gasoline yield (43.8 wt%) and lowest coke (4.7 wt%). The total aromatics yield from algae oil cracking is improved 7.8 wt% when MFI is loaded with gallium.

  3. The Future of Low-Carbon Transportation Fuels

    NASA Astrophysics Data System (ADS)

    Yang, Christopher; Yeh, Sonia

    2011-11-01

    Petroleum fuel uses make up essentially all of transportation fuel usage today and will continue to dominate transportation fuel usage well into future without any major policy changes. This chapter focuses on low-carbon transportation fuels, specifically, biofuels, electricity and hydrogen, that are emerging options to displace petroleum based fuels. The transition to cleaner, lower carbon fuel sources will need significant technology advancement, and sustained coordination efforts among the vehicle and fuel industry and policymakers/regulators over long period of time in order to overcome market barriers, consumer acceptance, and externalities of imported oil. We discuss the unique infrastructure challenges, and compare resource, technology, economics and transitional issues for each of these fuels. While each fuel type has important technical and implementation challenges to overcome (including vehicle technologies) in order to contribute a large fraction of our total fuel demand, it is important to note that a portfolio approach will give us the best chance of meeting stringent environmental and energy security goals for a sustainable transportation future.

  4. EVermont Renewable Hydrogen Production and Transportation Fueling System

    SciTech Connect

    Garabedian, Harold T. Wight, Gregory Dreier, Ken Borland, Nicholas

    2008-03-30

    A great deal of research funding is being devoted to the use of hydrogen for transportation fuel, particularly in the development of fuel cell vehicles. When this research bears fruit in the form of consumer-ready vehicles, will the fueling infrastructure be ready? Will the required fueling systems work in cold climates as well as they do in warm areas? Will we be sure that production of hydrogen as the energy carrier of choice for our transit system is the most energy efficient and environmentally friendly option? Will consumers understand this fuel and how to handle it? Those are questions addressed by the EVermont Wind to Wheels Hydrogen Project: Sustainable Transportation. The hydrogen fueling infrastructure consists of three primary subcomponents: a hydrogen generator (electrolyzer), a compression and storage system, and a dispenser. The generated fuel is then used to provide transportation as a motor fuel. EVermont Inc., started in 1993 by then governor Howard Dean, is a public-private partnership of entities interested in documenting and advancing the performance of advanced technology vehicles that are sustainable and less burdensome on the environment, especially in areas of cold climates, hilly terrain and with rural settlement patterns. EVermont has developed a demonstration wind powered hydrogen fuel producing filling system that uses electrolysis, compression to 5000 psi and a hydrogen burning vehicle that functions reliably in cold climates. And that fuel is then used to meet transportation needs in a hybrid electric vehicle whose internal combustion engine has been converted to operate on hydrogen Sponsored by the DOE EERE Hydrogen, Fuel Cells & Infrastructure Technologies (HFC&IT) Program, the purpose of the project is to test the viability of sustainably produced hydrogen for use as a transportation fuel in a cold climate with hilly terrain and rural settlement patterns. Specifically, the project addresses the challenge of building a renewable

  5. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications. [Radiation dose rates from shielded spent fuels and high-level radioactive waste

    SciTech Connect

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs.

  6. PLATINUM, FUEL CELLS, AND FUTURE ROAD TRANSPORT

    EPA Science Inventory

    A vehicle powered by a fuel cell will emit virtually no air polution and, depending on fuel choice, can substantially improve fuel economy above that of current technology. Those attributes are complementary to issues of increasing national importance including the effects of tra...

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

    SciTech Connect

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

    1982-11-01

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

  8. US Department of Energy fuel cell program for transportation applications

    NASA Astrophysics Data System (ADS)

    Patil, Pandit G.

    1992-01-01

    Fuel cells of offer promise as the best future replacement for internal combustion engines in transportation applications. Fuel cells operate more efficiently than internal combustion engines, and are capable of running on non-petroleum fuels such as methanol, ethanol, natural gas or hydrogen. Fuel cells can also have a major impact on improving air quality. They virtually eliminate particulates, NO(x) and sulfur oxide emissions, and significantly reduce hydrocarbons and carbon monoxide. The U.S. Department of Energy program on fuel cells for transportation applications is structured to advance fuel cells technologies from the R&D phase, through engineering design and scale-tip, to demonstration in cars, trucks, buses and locomotives, in order to provide energy savings, fuel flexibility and air quality improvements. This paper describes the present status of the U.S. program.

  9. Liquid-fueled SOFC power sources for transportation

    NASA Astrophysics Data System (ADS)

    Myles, K. M.; Doshi, R.; Kumar, R.; Krumpelt, M.

    Traditionally, fuel cells have been developed for space or stationary terrestrial applications. As the first commercial 200-kW systems were being introduced by ONSI and Fuji Electric, the potentially much larger, but also more challenging, application in transportation was beginning to be addressed. As a result, fuel cell-powered buses have been designed and built, and R&D programs for fuel cell-powered passenger cars have been initiated. The engineering challenge of eventually replacing the internal combustion engine in buses, trucks, and passenger cars with fuel cell systems is to achieve much higher power densities and much lower costs than obtainable in systems designed for stationary applications. At present, the leading fuel cell candidate for transportation applications is, without question, the polymer electrolyte fuel cell (PEFC). Offering ambient temperature start-up and the potential for a relatively high power density, the polymer technology has attracted the interest of automotive manufacturers worldwide. But the difficulties of fuel handling for the PEFC have led to a growing interest in exploring the prospects for solid oxide fuel cells (SOFCs) operating on liquid fuels for transportation applications. Solid oxide fuel cells are much more compatible with liquid fuels (methanol or other hydrocarbons) and are potentially capable of power densities high enough for vehicular use. Two SOFC options for such use are discussed in this report.

  10. Brazil's program to use ethanol as transportation fuel loses steam

    SciTech Connect

    Anderson, E.V.

    1993-10-18

    The spectacular growth of Proalcool, Brazil's vaunted program to pump sugar-based ethanol into the country's transportation fuel market, has ground to a virtual halt. Ethanol's share of Brazil's transportation fuel market slipped from its peak of 28% in 1989 to less than 25% in 1991. During that same time, gasoline recovered some of its lost share, while diesel remains the preeminent fuel in Brazil. Proalcool still is by far the largest alternative fuel program in the world. But Proalcool has fallen victim to its own success, Brazil's flattering economy and rampant inflation, and the idiosyncrasies of the country's refinery economics.

  11. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect

    Durbin, Samuel; Lindgren, Eric R.

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  12. Minimum criticality dose evaluation for the Irradiated Fuel Storage Facility

    SciTech Connect

    Kim, S.S.

    1999-09-01

    The Irradiated Fuel Storage Facility (IFSF) is a government-owned, contractor-operated facility located at the Idaho National Engineering and Environmental Laboratory within the Idaho Nuclear Technology and Engineering Center. The mission of the facility is to provide safe dry storage for various types of irradiated fuels. Included are fuel elements such as irradiated ATR, EBR, MTR, Fort St. Vrain, TRIGA, and ROVER Parka fuels. Fuels requiring dry storage are received at the IFSF in fuel-shipping casks. At the facility receiving dock, the casks are removed from the transport vehicle, positioned in a cask transport car, and moved into the fuel-handling cave. Several functions are performed in the fuel-handling cave, including transferring fuel from shipping casks to storage canisters, preparing fuel elements for storage and processing. The minimum postulated criticality dose calculations were performed for the cask-receiving and fuel-handling areas to place criticality alarm system (CAS) detectors. The number of fissions for the minimum accident of concern is based on a dose of 20-rad air at 2 m in 1 min. The eigenvalue calculations were first performed to determine the size of the critical source. Then, two sets of fixed-source calculations were followed to calculate contributions from neutron and capture gamma rays and from prompt gamma rays. Two sets of MCNP calculations involved point and spherical critical sources. Validity of the Monte Carlo results was tested against ANISN deterministic calculations. The flux-to-dose conversion factors are based on ANSI/ANS-6.1.1-1977. All of the MCNP runs used continuous-energy ENDF/B-V cross sections. The BUGLE-80 cross-section library was used for the ANISN calculations.

  13. Reexamination of spent fuel shipment risk estimates

    SciTech Connect

    COOK,J.R.; SPRUNG,JEREMY L.

    2000-04-25

    The risks associated with the transport of spent nuclear fuel by truck and rail have been reexamined and compared to results published in NUREG-O170 and the Modal Study. The full reexamination considered transport of PWR and BWR spent fuel by truck and rail in four generic Type B spent fuel casks. Because they are typical, this paper presents results only for transport of PWR spent fuel in steel-lead steel casks. Cask and spent fuel response to collision impacts and fires were evaluated by performing three-dimensional finite element and one-dimensional heat transport calculations. Accident release fractions were developed by critical review of literature data. Accident severity fractions were developed from Modal Study truck and rail accident event trees, modified to reflect the frequency of occurrence of hard and soft rock wayside route surfaces as determined by analysis of geographic data. Incident-free population doses and the population dose risks associated with the accidents that might occur during transport were calculated using the RADTRAN 5 transportation risk code. The calculated incident-free doses were compared to those published in NUREG-O170. The calculated accident dose risks were compared to dose risks calculated using NUREG-0170 and Modal Study accident source terms. The comparisons demonstrated that both of these studies made a number of very conservative assumptions about spent fuel and cask response to accident conditions, which caused their estimates of accident source terms, accident frequencies, and accident consequences to also be very conservative. The results of this study and the previous studies demonstrate that the risks associated with the shipment of spent fuel by truck or rail are very small.

  14. Transportation capabilities study of DOE-owned spent nuclear fuel

    SciTech Connect

    Clark, G.L.; Johnson, R.A.; Smith, R.W.; Abbott, D.G.; Tyacke, M.J.

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  15. 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.

  16. Hydrocracking of the oils of Botryococcus braunii for transport fuels

    SciTech Connect

    Hillen, L.W.; Pollard, G.; Wake, L.V.; White, N.

    1982-01-01

    Hydrocarbon oils of the alga Botryococcus braunii, extracted from a natural ''bloom'' of the plant, have been hydrocracked to produce a distillate comprising 67% gasoline fraction, 15% aviation turbine fuel fraction, 15% diesel fuel fraction, and 3% residual oil. The distillate was examined by a number of standard petroleum industry test methods. This preliminary investigation indicates that the oils of B. braunii are suitable as a feedstock material for hydrocracking to transportation fuels;

  17. Hydrocracking of the oils of Botryococcus braunii to transport fuels

    SciTech Connect

    Hillen, L.W.; Pollard, G.; Wake, L.V.; White, N.

    1980-07-01

    Hydrocarbon oils of the alga Botryococcus braunii, extracted from a natural 'bloom' of the plant, have been hydrocracked to produce a distillate comprising 67% a petrol fraction, 15% an aviation turbine fuel fraction, 15% a diesel fuel fraction and 3% residual oil. The distillate was examined by a number of standard petroleum industry test methods. This preliminary investigation indicates that the oils of B. braunii are suitable as a feedstock material for hydrocracking to transport fuels.

  18. METHANOL PRODUCTION FROM BIOMASS AND NATURAL GAS AS TRANSPORTATION FUEL

    EPA Science Inventory

    Two processes are examined for production of methanol. They are assessed against the essential requirements of a future alternative fuel for road transport: that it (i) is producible in amounts comparable to the 19 EJ of motor fuel annually consumed in the U.S., (ii) minimizes em...

  19. On direct and indirect methanol fuel cells for transportation applications

    SciTech Connect

    Gottesfield, S.

    1996-04-01

    Research on direct oxidation methanol fuel cells (DMFCs) and polymer electrolyte fuel cells (PEFCs) is discussed. Systems considered for transportation applications are addressed. The use of platinum/ruthenium anode electrocatalysts and platinum cathode electrocatalysts in polymer electrolyte DMFCs has resulted in significant performance enhancements.

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

    SciTech Connect

    S. G. Johnson; K. L. Lively

    2010-05-01

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

  1. INL Site Executable Plan for Energy and Transportation Fuels Management

    SciTech Connect

    Ernest L. Fossum

    2008-11-01

    It is the policy of the Department of Energy (DOE) that sustainable energy and transportation fuels management will be integrated into DOE operations to meet obligations under Executive Order (EO) 13423 "Strengthening Federal Environmental, Energy, and Transportation Management," the Instructions for Implementation of EO 13423, as well as Guidance Documents issued in accordance thereto and any modifcations or amendments that may be issued from time to time. In furtherance of this obligation, DOE established strategic performance-based energy and transportation fuels goals and strategies through the Transformational Energy Action Management (TEAM) Initiative, which were incorporated into DOE Order 430.2B "Departmental Energy, Renewable energy, and Transportation Management" and were also identified in DOE Order 450.1A, "Environmental Protection Program." These goals and accompanying strategies are to be implemented by DOE sites through the integration of energy and transportation fuels management into site Environmental Management Systems (EMS).

  2. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    SciTech Connect

    Yuan, Y.C.; Chen, S.Y.; Biwer, B.M.; LePoire, D.J.

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

  3. Environmental benefits of transport and stationary fuel cells

    NASA Astrophysics Data System (ADS)

    Hart, David; Hörmandinger, Günter

    The potential environmental benefits of using fuel cells in cars, buses and stationary combined heat and power (CHP) plants of different sizes have not been well-researched. This environmental analysis was conducted for the UK on a `full fuel cycle' basis, encompassing all greenhouse gas and regulated pollutant emissions for the supply chain and end-use technology under consideration. Solid polymer fuel cells (SPFCs) with methanol or natural gas reformers were analysed for cars, SPFCs and phosphoric acid fuel cells (PAFCs) with on-board hydrogen for buses. CHP plants were PAFCs or solid oxide fuel cells (SOFCs). Each option was compared with one or more conventional technologies. In all cases fuel cell technologies have substantially reduced emissions in comparison with conventional technologies. Regulated emissions are lowest, by up to two orders of magnitude, and those that do occur are primarily in the fuel supply chain. The fuel cell technologies are more efficient in all cases, and carbon dioxide (CO2) emissions are reduced broadly in line with energy savings. Methane emissions increase due to fuel switching, e.g. from petrol to natural gas powered buses, but from a very low base. The study pinpoints some areas in which alternative approaches could be made - the methods for generating and transporting hydrogen have a significant bearing on energy consumption and emissions. However, it is clear that from an overall emissions perspective the use of fuel cells in transport and power generation is highly beneficial.

  4. Fuel conservation possibilities for terminal area compatible transport aircraft

    NASA Technical Reports Server (NTRS)

    Hanks, G. W.; Heath, A. R., Jr.

    1975-01-01

    Design characteristics that would reduce mission fuel consumption and improve terminal-area operations for advanced transports are discussed. Sensitivity studies of the effects of cruise speed, wing geometry, propulsion cycle, operational procedures, and payload on fuel usage are presented and utilized to arrive at a conceptual configuration which offers mission fuel savings as well as desirable operational characteristics in the terminal area. Technical and economic evaluation is provided in the form of a comparison of the resulting configuration with transports reflecting the current level of technology. The research and technology programs required to realize potential benefits are described.

  5. Barriers to the utilization of synthetic fuels for transportation

    NASA Technical Reports Server (NTRS)

    Parker, H. W.; Reilly, M. J.

    1981-01-01

    The principal types of engines for transportation uses are reviewed and the specifications for conventional fuels are compared with specifications for synthetic fuels. Synfuel processes nearing the commercialization phase are reviewed. The barriers to using synfuels can be classified into four groups: technical, such as the uncertainty that a new engine design can satisfy the desired performance criteria; environmental, such as the risk that the engine emissions cannot meet the applicable environmental standards; economic, including the cost of using a synfuel relative to conventional transportation fuels; and market, involving market penetration by offering new engines, establishing new distribution systems and/or changing user expectations.

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

    SciTech Connect

    C. Sanders; D. Kimball

    2005-08-02

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

  7. Multi-fuel reformers for fuel cells used in transportation. Multi-fuel reformers: Phase 1 -- Final report

    SciTech Connect

    Not Available

    1994-05-01

    DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

  8. 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.

  9. 30 CFR 75.1906 - Transport of diesel fuel.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... transporting safety cans containing diesel fuel must have at least two multipurpose, dry chemical type (ABC) fire extinguishers, listed or approved by a nationally recognized independent testing laboratory and... HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Diesel-Powered Equipment § 75.1906 Transport...

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

    SciTech Connect

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

    1992-04-01

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

  11. An assessment and comparison of fuel cells for transportation applications

    NASA Astrophysics Data System (ADS)

    Krumpelt, M.; Christianson, C. C.

    1989-09-01

    Fuel cells offer the potential of a clean, efficient power source for buses, cars, and other transportation applications. When the fuel cell is run on methanol, refueling would be as rapid as with gasoline-powered internal combustion engines, providing a virtually unlimited range while still maintaining the smooth and quiet acceleration that is typical for electric vehicles. The advantages and disadvantages of five types of fuel cells are reviewed and analyzed for a transportation application: alkaline, phosphoric acid, proton exchange membrane, molten carbonate, and solid oxide. The status of each technology is discussed, system designs are reviewed, and preliminary comparisons of power densities, start-up times, and dynamic response capabilities are made. To test the concept, a fuel cell/battery powered urban bus appears to be a good first step that can be realized today with phosphoric acid cells. In the longer term, the proton exchange membrane and solid oxide fuel cells appear to be superior.

  12. Study of methane fuel for subsonic transport aircraft

    NASA Technical Reports Server (NTRS)

    Carson, L. K.; Davis, G. W.; Versaw, E. F.; Cunnington, G. R., Jr.; Daniels, E. J.

    1980-01-01

    The cost and performance were defined for commercial transport using liquid methane including its fuel system and the ground facility complex required for the processing and storage of methane. A cost and performance comparison was made with Jet A and hydrogen powered aircraft of the same payload and range capability. Extensive design work was done on cryogenic fuel tanks, insulation systems as well as the fuel system itself. Three candidate fuel tank locations were evaluated, i.e., fuselage tanks, wing tanks or external pylon tanks.

  13. Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

    SciTech Connect

    M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

    2007-10-01

    The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department of Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor

  14. Site and facility waste transport planning documents (SPDs) status and findings

    SciTech Connect

    Schmid, S P; Danese, F L; Wankerl, M W

    1993-05-01

    Site and Facility Waste Transportation Services Planning Documents (SPDS) are detailed desk-top reference documents that initiate planning for shipping commercial spent nuclear fuel (SNF) from the sites where it is currently generated and/or stored to another location. Because of the unique design features and individual variations in the spent fuel handling and cask loading operation requirements for each facility, one SPD will be written for each of the commercial facilities currently expected to deliver SNF into the Civilian Radioactive Waste Management System (CRWMS) disposal system. One primary purpose of an SPD is to initiate the discussions that will lead to a determination of the type of spent fuel cask and transport mode that will be used to transport spent fuel from each facility. The initial assessment of which cask type and mode would best serve each facility is based on the principle that the largest possible capacity casks should be used at the greatest number of facilities to reduce the total number of spent fuel shipments. The final selection of cask and transportation mode will be arrived at following discussion with the facility licensed operator. Once agreed upon by OCRWM and the facility owner, the SPD wig be used as a primary input to the development of a Site Specific Servicing Plan (SSSP) that will detail chosen servicing options for the individual site. This paper will discuss the purpose and development of SPDs and the preliminary results of an evaluation of the ability of delivering facilities to handle and ship spent fuel casks within the confines of the local nation infrastructure.

  15. Proton exchange membrane fuel cell technology for transportation applications

    SciTech Connect

    Swathirajan, S.

    1996-04-01

    Proton Exchange Membrane (PEM) fuel cells are extremely promising as future power plants in the transportation sector to achieve an increase in energy efficiency and eliminate environmental pollution due to vehicles. GM is currently involved in a multiphase program with the US Department of Energy for developing a proof-of-concept hybrid vehicle based on a PEM fuel cell power plant and a methanol fuel processor. Other participants in the program are Los Alamos National Labs, Dow Chemical Co., Ballard Power Systems and DuPont Co., In the just completed phase 1 of the program, a 10 kW PEM fuel cell power plant was built and tested to demonstrate the feasibility of integrating a methanol fuel processor with a PEM fuel cell stack. However, the fuel cell power plant must overcome stiff technical and economic challenges before it can be commercialized for light duty vehicle applications. Progress achieved in phase I on the use of monolithic catalyst reactors in the fuel processor, managing CO impurity in the fuel cell stack, low-cost electrode-membrane assembles, and on the integration of the fuel processor with a Ballard PEM fuel cell stack will be presented.

  16. Ultraclean Fuels Production and Utilization for the Twenty-First Century: Advances toward Sustainable Transportation Fuels

    SciTech Connect

    Fox, Elise B.; Liu, Zhong-Wen; Liu, Zhao-Tie

    2013-11-21

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  17. High speed commercial transport fuels considerations and research needs

    NASA Technical Reports Server (NTRS)

    Lee, C. M.; Niedzwiecki, R. W.

    1989-01-01

    NASA is currently evaluating the potential of incorporating High Speed Civil Transport (HSCT) aircraft in the commercial fleet in the beginning of the 21st century. NASA sponsored HSCT enabling studies currently underway with airframers and engine manufacturers, are addressing a broad range of technical, environmental, economic, and related issues. Supersonic cruise speeds for these aircraft were originally focused in the Mach 2 to 5 range. At these flight speeds, both jet fuels and liquid methane were considered potential fuel candidates. For the year 2000 to 2010, cruise Mach numbers of 2 to 3+ are projected for aircraft fuel with thermally stable liquid jet fuels. For 2015 and beyond, liquid methane fueled aircraft cruising at Mach numbers of 4+ may be viable candidates. Operation at supersonic speeds will be much more severe than those encountered at subsonic flight. One of the most critical problems is the potential deterioration of the fuel due to the high temperature environment. HSCT fuels will not only be required to provide the energy necessary for flight, but will also be subject to aerodynamic heating and, will be required to serve as the primary heat sink for cooling the engine and airframe. To define fuel problems for high speed flight, a fuels workshop was conducted at NASA Lewis Research Center. The purpose of the workshop was to gather experts on aviation fuels, airframe fuel systems, airport infrastructure, and combustion systems to discuss high speed fuel alternatives, fuel supply scenarios, increased thermal stability approaches and measurements, safety considerations, and to provide directional guidance for future R and D efforts. Subsequent follow-up studies defined airport infrastructure impacts of high speed fuel candidates. The results of these activities are summarized. In addition, an initial case study using modified in-house refinery simulation model Gordian code (1) is briefly discussed. This code can be used to simulate different

  18. Reimagining liquid transportation fuels : sunshine to petrol.

    SciTech Connect

    Johnson, Terry Alan; Hogan, Roy E., Jr.; McDaniel, Anthony H.; Siegel, Nathan Phillip; Dedrick, Daniel E.; Stechel, Ellen Beth; Diver, Richard B., Jr.; Miller, James Edward; Allendorf, Mark D.; Ambrosini, Andrea; Coker, Eric Nicholas; Staiger, Chad Lynn; Chen, Ken Shuang; Ermanoski, Ivan; Kellog, Gary L.

    2012-01-01

    Two of the most daunting problems facing humankind in the twenty-first century are energy security and climate change. This report summarizes work accomplished towards addressing these problems through the execution of a Grand Challenge LDRD project (FY09-11). The vision of Sunshine to Petrol is captured in one deceptively simple chemical equation: Solar Energy + xCO{sub 2} + (x+1)H{sub 2}O {yields} C{sub x}H{sub 2x+2}(liquid fuel) + (1.5x+.5)O{sub 2} Practical implementation of this equation may seem far-fetched, since it effectively describes the use of solar energy to reverse combustion. However, it is also representative of the photosynthetic processes responsible for much of life on earth and, as such, summarizes the biomass approach to fuels production. It is our contention that an alternative approach, one that is not limited by efficiency of photosynthesis and more directly leads to a liquid fuel, is desirable. The development of a process that efficiently, cost effectively, and sustainably reenergizes thermodynamically spent feedstocks to create reactive fuel intermediates would be an unparalleled achievement and is the key challenge that must be surmounted to solve the intertwined problems of accelerating energy demand and climate change. We proposed that the direct thermochemical conversion of CO{sub 2} and H{sub 2}O to CO and H{sub 2}, which are the universal building blocks for synthetic fuels, serve as the basis for this revolutionary process. To realize this concept, we addressed complex chemical, materials science, and engineering problems associated with thermochemical heat engines and the crucial metal-oxide working-materials deployed therein. By project's end, we had demonstrated solar-driven conversion of CO{sub 2} to CO, a key energetic synthetic fuel intermediate, at 1.7% efficiency.

  19. Fuel-cycle greenhouse gas emissions impacts of alternative transportation fuels and advanced vehicle technologies.

    SciTech Connect

    Wang, M. Q.

    1998-12-16

    At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and fuel-cell vehicles can reduce GHG emissions by over 40%. Spark ignition engines and fuel-cell vehicles powered by cellulosic ethanol and solar hydrogen (for fuel-cell vehicles only) can reduce GHG emissions by over 80%. In conclusion, both near- and long-term alternative fuels and advanced transportation technologies can play a role in reducing the United States GHG emissions.

  20. Bioethanol: A Renewable Transportation Fuel from Biomass

    SciTech Connect

    Riley, C. J.

    2002-01-01

    Environmentally acceptable scenarios exist in which bioethanol is a major energy carrier for a sustainable transportation sector. Significant progress has been made in developing the new technologies needed but they remain to be proven at the commercial scale.

  1. Modelling of RBMK-1500 SNF storage casks activation during very long term storage.

    PubMed

    Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Ragaisis, Valdas

    2016-09-01

    Existing interim spent nuclear fuel storage facility (SNFSF) at Ignalina nuclear power plant in Lithuania is fully loaded with CASTOR(®)RBMK-1500 and CONSTOR(®)RBMK-1500 storage casks. The planned lifetime of these casks is 50 years and the first loaded cask was moved to the SNFSF in 1999. The start of operation of disposal facility in Lithuania is foreseen later than the planned interim storage ends. So, the possibilities to extend the storage period over 50 years should be considered. Therefore, the casks decommissioning issues should be taken into account, as due to prolonged neutron irradiation casks materials could became activated. This paper presents modelling results of storage casks neutron activation during 300 year storage period. Modelling results show, that after 50 years of storage, side-wall and bottom of CASTOR(®)RBMK-1500 cask are activated above clearance criteria. However, for 100-300 year storage period all of the casks components could be free released. PMID:27344524

  2. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Mertyurek, Ugur; Radulescu, Georgeta

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  3. Synthetic fuels for ground transportation with special emphasis on hydrogen

    NASA Technical Reports Server (NTRS)

    Singh, J. J.

    1975-01-01

    The role of various synthetic fuels, for ground transportation in the United States, was examined for the near term (by 1985) and the longer term applications (1985-2000 and beyond 2000). Feasible options include synthetic oil, methanol, electric propulsion, and hydrogen. It is concluded that (1) the competition during the next 50 years will be for the fuels of all types, rather than among the fuels; (2) extensive domestic oil and gas exploration should be initiated concurrent with the development of several alternate fuels and related ancillaries; and (3) hydrogen, as an automotive fuel, seems to be equivalent to gasoline for optimum fuel to air mixtures. As a pollution free, high energy density fuel, hydrogen deserves consideration as the logical replacement for the hydrocarbons. Several research and development requirements, essential for the implementation of hydrogen economy for ground transportation, were identified. Extensive engineering development and testing activities should be initiated to establish hydrogen as the future automotive fuel, followed by demonstration projects and concerted efforts at public education.

  4. CONVERTING PYROLYSIS OILS TO RENEWABLE TRANSPORT FUELS: PROCESSING CHALLENGES & OPPORTUNITIES

    SciTech Connect

    Holmgren, Jennifer; Nair, Prabhakar N.; Elliott, Douglas C.; Bain, Richard; Marinangelli, Richard

    2008-03-11

    To enable a sustained supply of biomass-based transportation fuels, the capability to process feedstocks outside the food chain must be developed. Significant industry efforts are underway to develop these new technologies, such as converting cellulosic wastes to ethanol. UOP, in partnership with U.S. Government labs, NREL and PNNL, is developing an alternate route using cellulosic feedstocks. The waste biomass is first subjected to a fast pyrolysis operation to generate pyrolysis oil (pyoil for short). Current efforts are focused on developing a thermochemical platform to convert pyoils to renewable gasoline, diesel and jet fuel. The fuels produced will be indistinguishable from their fossil fuel counterparts and, therefore, will be compatible with existing transport and distribution infrastructure.

  5. Moving beyond alternative fuel hype to decarbonize transportation

    NASA Astrophysics Data System (ADS)

    Melton, Noel; Axsen, Jonn; Sperling, Daniel

    2016-03-01

    In the past three decades, government, industry and other stakeholders have repeatedly been swept up with the ‘fuel du jour’, claiming that a particular alternative fuel vehicle (AFV) technology can succeed in replacing conventional gasoline-powered vehicles. However, AFV technologies have experienced relatively little success, with fossil fuels still accounting for about 95% of global transport energy use. Here, using the US as a case study, we conduct a media analysis to show how society’s attention has skipped among AFV types between 1980 and 2013, including methanol, natural gas, plug-in electric, hybrid electric, hydrogen and biofuels. Although our results provide no indication as to whether hype ultimately has a net positive or negative impact on AFV innovation, we offer several recommendations that governments can follow to move past hype to support significant AFV adoption and displace fossil fuel use in the transportation sector.

  6. Spent nuclear fuel removal program at the West Valley Demonstration Project: Topical report

    SciTech Connect

    Connors, B. J.; Golden, M. P.; Valenti, P. J.; Winkel, J. J.

    1987-03-01

    The spent nuclear fuel removal program at the West Valley Demonstration Project (WVDP) consisted of removing the spent nuclear fuel (SNF) assemblies from the storage pool in the plant, loading them in shielded casks, and preparing the casks for transportation. So far, four fuel removal campaigns have been completed with the return of 625 spent nuclear fuel assemblies to their four utility owners. A fifth campaign, which is not yet completed, will transfer the remaining 125 fuel assemblies to a government site in Idaho. A spent fuel rod consolidation demonstration has been completed, and the storage canisters and their racks are being removed from the fuel receiving and storage pool to make way for installation of the size reduction equipment. A brief history of the West Valley reprocessing plant and the events leading to the storage and ownership of the spent nuclear fuel assemblies and their subsequent removal from West Valley are also recorded as background information. 3 refs., 16 figs., 9 tabs.

  7. Fuel consumption improvement in current transport engines

    NASA Technical Reports Server (NTRS)

    Hines, R. W.; Ziemianski, J. A.

    1978-01-01

    A review is conducted of improvements which can be made with respect to the fuel consumption of current engines and new production versions of current engines. A description is presented of an engine diagnostics program which has the objective to identify and quantify the causes and sources of performance deterioration in the JT9D turbofan engine and to develop basic data which will be applied to minimize performance degradation of current and future engines. General areas where performance losses occur are examined, taking into account seals, blades and vanes, and cases. Potential performance improvement concepts are related to improved component aerodynamics, improved flowpath sealing, blade tip clearance control, improved turbine cooling effectiveness, improved turbine materials and coatings, duct and nozzle aerodynamic refinements, nacelle aerodynamic refinements, forced exhaust mixers, advanced nacelle materials, and advanced fuel control.

  8. Solar Energy for Transportation Fuel (LBNL Science at the Theater)

    ScienceCinema

    Lewis, Nate

    2011-04-28

    Nate Lewis' talk looks at the challenge of capturing solar energy and storing it as an affordable transportation fuel - all on a scale necessary to reduce global warming. Overcoming this challenge will require developing new materials that can use abundant and inexpensive elements rather than costly and rare materials. He discusses the promise of new materials in the development of carbon-free alternatives to fossil fuel.

  9. Hydrocracking of the oils of Botryococcus braunii to transport fuels

    SciTech Connect

    Hillen, L.W.; Pollard, G.; Wakr, L.V.; White, N.

    1982-01-01

    Hydrocarbon oils of the alga Botryococcus braunii, extracted from a natural ''bloom'' of the plant, have been hydrocracked to produce a distillate comprising 67% gasoline fraction, 15% aviation turbine fuel fraction, 15% diesel fule fraction, and 3% residual oil. The distillate was examined by a number of standard petroleum industry test methods. This preliminary investigation indicates that the oils of B. braunii are suitable as a feedstock material for hydrocracking to transport fuels.

  10. Solar Energy for Transportation Fuel (LBNL Science at the Theater)

    SciTech Connect

    Lewis, Nate

    2008-05-12

    Nate Lewis' talk looks at the challenge of capturing solar energy and storing it as an affordable transportation fuel - all on a scale necessary to reduce global warming. Overcoming this challenge will require developing new materials that can use abundant and inexpensive elements rather than costly and rare materials. He discusses the promise of new materials in the development of carbon-free alternatives to fossil fuel.

  11. 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).

  12. A smooth transition to hydrogen transportation fuel

    SciTech Connect

    Berry, G.D.; Smith, J.R.; Schock, R.N.

    1995-04-14

    The goal of this work is to examine viable near-term infrastructure options for a transition to hydrogen fueled vehicles and to suggest profitable directions for technology development. The authors have focused in particular on the contrasting options of decentralized production using the existing energy distribution network, and centralized production of hydrogen with a large-scale infrastructure. Delivered costs have been estimated using best available industry cost and deliberately conservative economic assumptions. The sensitivities of these costs have then been examined for three small-scale scenarios: (1) electrolysis at the home for one car, and production at the small station scale (300 cars/day), (2) conventional alkaline electrolysis and (3) steam reforming of natural gas. All scenarios assume fueling a 300 mile range vehicle with 3.75 kg. They conclude that a transition appears plausible, using existing energy distribution systems, with home electrolysis providing fuel costing 7.5 to 10.5{cents}/mile, station electrolysis 4.7 to 7.1{cents}/mile, and steam reforming 3.7 to 4.7{cents}/mile. The average car today costs about 6{cents}/mile to fuel. Furthermore, analysis of liquid hydrogen delivered locally by truck from central processing plants can also be competitive at costs as low as 4{cents}/mile. These delivered costs are equal to $30 to $70 per GJ, LHV. Preliminary analysis indicates that electricity transmission costs favor this method of distributing energy, until very large (10 GW) hydrogen pipelines are installed. This indicates that significant hydrogen pipeline distribution will be established only when significant markets have developed.

  13. Safety analysis of irradiated nuclear fuel transportation container

    SciTech Connect

    Uspuras, E.; Rimkevicius, S.

    2007-07-01

    Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit I on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level. (authors)

  14. Fuel conservation merits of advanced turboprop transport aircraft

    NASA Technical Reports Server (NTRS)

    Revell, J. D.; Tullis, R. H.

    1977-01-01

    The advantages of a propfan powered aircraft for the commercial air transportation system were assessed by the comparison with an equivalent turbofan transport. Comparisons were accomplished on the basis of fuel utilization and operating costs, as well as aircraft weight and size. Advantages of the propfan aircraft, concerning fuel utilization and operating costs, were accomplished by considering: (1) incorporation of propfan performance and acoustic data; (2) revised mission profiles (longer design range and reduction in; and cruise speed) (3) utilization of alternate and advanced technology engines.

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

    SciTech Connect

    Einziger, R.E.; Compton, K.; Gordon, M.; Ahn, T.; Gonzales, H.; Pan, Y.

    2012-07-01

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

  16. HYDROGEN COMMERCIALIZATION: TRANSPORTATION FUEL FOR THE 21ST CENTURY

    SciTech Connect

    APOLONIO DEL TORO

    2008-05-27

    Since 1999, SunLine Transit Agency has worked with the U.S. Department of Energy (DOE), U.S. Department of Defense (DOD), and the U.S. Department of Transportation (DOT) to develop and test hydrogen infrastructure, fuel cell buses, a heavy-duty fuel cell truck, a fuel cell neighborhood electric vehicle, fuel cell golf carts and internal combustion engine buses operating on a mixture of hydrogen and compressed natural gas (CNG). SunLine has cultivated a rich history of testing and demonstrating equipment for leading industry manufacturers in a pre-commercial environment. Visitors to SunLine's "Clean Fuels Mall" from around the world have included government delegations and agencies, international journalists and media, industry leaders and experts and environmental and educational groups.

  17. Thermoelectric Powered Wireless Sensors for Dry-Cask Storage

    NASA Astrophysics Data System (ADS)

    Carstens, Thomas Alan

    This study focuses on the development of self-powered wireless sensors. These sensors can be used to measure key parameters in extreme environments; e.g., temperature monitoring for spent nuclear fuel during dry-cask storage. This study has developed a design methodology for these self-powered monitoring systems. The main elements that constitute this work consist of selecting and testing a power source for the wireless sensor, determination of the attenuation of the wireless signal, and testing the wireless sensor circuitry in an extreme environment. OrigenArp determined the decay heat and gamma/neutron source strength of the spent fuel throughout the service life of the dry-cask. A first principles analysis modeled the temperatures inside the dry-cask. A finite-element heat transfer code calculated the temperature distribution of the thermoelectric and heat sink. The temperature distributions determine the power produced by the thermoelectric. It was experimentally verified that a thermoelectric generator (HZ-14) with a DC/DC converter (Linear Technology LTC3108EDE) can power a transceiver (EmbedRF) at condition which represent prototypical conditions throughout and beyond the service life of the dry-cask. The wireless sensor is required to broadcast with enough power to overcome the attenuation from the dry-cask. It will be important to minimize the attenuation of the signal in order to broadcast with a small transmission power. To investigate the signal transmission through the dry-cask, CST Microwave Studio was used to determine the scattering parameter S2,1 for a horizontal dry-cask. Important parameters that can influence the transmission of the signal are antenna orientation, antenna placement, and transmission frequency. The thermoelectric generator, DC/DC converter, and transceiver were exposed to 60Co gamma radiation (exposure rate170.3 Rad/min) at the University of Wisconsin Medical Radiation Research Center. The effects of gamma radiation on the

  18. 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.

  19. Actinide partitioning-transmutation program final report. IV. Miscellaneous aspects. [Transport; fuel fabrication; decay; policy; economics

    SciTech Connect

    Alexander, C.W.; Croff, A.G.

    1980-09-01

    This report discusses seven aspects of actinide partitioning-transmutation (P-T) which are important in any complete evaluation of this waste treatment option but which do not fall within other major topical areas concerning P-T. The so-called miscellaneous aspects considered are (1) the conceptual design of a shipping cask for highly neutron-active fresh and spent P-T fuels, (2) the possible impacts of P-T on mixed-oxide fuel fabrication, (3) alternatives for handling the existing and to-be-produced spent fuel and/or wastes until implementation of P-T, (4) the decay and dose characteristics of P-T and standard reactor fuels, (5) the implications of P-T on currently existing nuclear policy in the United States, (6) the summary costs of P-T, and (7) methods for comparing the risks, costs, and benefits of P-T.

  20. Energy demand analysis and alternative fuels. Transportation research record

    SciTech Connect

    Dingemans, D.; Sperling, D.; Greene, D.L.; Hu, P.S.; Hallet, P.

    1986-01-01

    Contents include: Mental maps and the refueling behavior of vehicle drivers; A functional form analysis of the short-run demand for travel and gasoline by one-vehicle households; An assessment methodology for alternative fuels technologies; Drive-up windows, energy, and air quality; Travel characteristics and transportation energy consumption patterns of minority and poor households; An investigation into the use of market segmentation analysis for transportation energy planning.

  1. 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.

  2. RECENT TRENDS IN EMERGING TRANSPORTATION FUELS AND ENERGY CONSUMPTION

    SciTech Connect

    Bunting, Bruce G

    2012-01-01

    Abundance of energy can be improved both by developing new sources of fuel and by improving efficiency of energy utilization, although we really need to pursue both paths to improve energy accessibility in the future. Currently, 2.7 billion people or 38% of the world s population do not have access to modern cooking fuel and depend on wood or dung and 1.4 billion people or 20% do not have access to electricity. It is estimated that correcting these deficiencies will require an investment of $36 billion dollars annually through 2030. In growing economies, energy use and economic growth are strongly linked, but energy use generally grows at a lower rate due to increased access to modern fuels and adaptation of modern, more efficient technology. Reducing environmental impacts of increased energy consumption such as global warming or regional emissions will require improved technology, renewable fuels, and CO2 reuse or sequestration. The increase in energy utilization will probably result in increased transportation fuel diversity as fuels are shaped by availability of local resources, world trade, and governmental, environmental, and economic policies. The purpose of this paper is to outline some of the recently emerging trends, but not to suggest winners. This paper will focus on liquid transportation fuels, which provide the highest energy density and best match with existing vehicles and infrastructure. Data is taken from a variety of US, European, and other sources without an attempt to normalize or combine the various data sources. Liquid transportation fuels can be derived from conventional hydrocarbon resources (crude oil), unconventional hydrocarbon resources (oil sands or oil shale), and biological feedstocks through a variety of biochemical or thermo chemical processes, or by converting natural gas or coal to liquids.

  3. Off-Highway Transportation-Related Fuel Use

    SciTech Connect

    Davis, S.C.

    2004-05-08

    The transportation sector includes many subcategories--for example, on-highway, off-highway, and non-highway. Use of fuel for off-highway purposes is not well documented, nor is the number of off-highway vehicles. The number of and fuel usage for on-highway and aviation, marine, and rail categories are much better documented than for off-highway land-based use. Several sources document off-highway fuel use under specific conditions--such as use by application (e.g., recreation) or by fuel type (e.g., gasoline). There is, however, no single source that documents the total fuel used off-highway and the number of vehicles that use the fuel. This report estimates the fuel usage and number of vehicles/equipment for the off-highway category. No new data have been collected nor new models developed to estimate the off-highway data--this study is limited in scope to using data that already exist. In this report, unless they are being quoted from a source that uses different terminology, the terms are used as listed below. (1) ''On-highway/on-road'' includes land-based transport used on the highway system or other paved roadways. (2) ''Off-highway/off-road'' includes land-based transport not using the highway system or other paved roadways. (3) ''Non-highway/non-road'' includes other modes not traveling on highways such as aviation, marine, and rail. It should be noted that the term ''transportation'' as used in this study is not typical. Generally, ''transportation'' is understood to mean the movement of people or goods from one point to another. Some of the off-highway equipment included in this study doesn't transport either people or goods, but it has utility in movement (e.g., a forklift or a lawn mower). Along these lines, a chain saw also has utility in movement, but it cannot transport itself (i.e., it must be carried) because it does not have wheels. Therefore, to estimate the transportation-related fuel used off-highway, transportation equipment is defined to

  4. 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.

  5. Specific transport and storage solutions : waste management facing current and future stakes of the nuclear fuel cycle

    SciTech Connect

    Choho, T.; Blachet, L.; Deniau, H.; Gagner, L.; Gendreau, F.; Presta, A.

    2007-07-01

    With major projects ongoing or being planned, and also with the daily management of radioactive waste from nuclear facilities, the role of transport and/or storage packaging has been often overlooked. Indeed, the packaging development process and transport solutions implemented are a key part of the waste management challenge : protection of people and of the environment. During over four decades, the AREVA Group has developed a complete and coherent system for the transport of waste produced by nuclear industries. The transport solutions integrate the factors to consider, as industrial transportation needs, various waste forms, associated hazards and current regulations. Thus, TN International has designed, licensed and manufactured a large number of different transport, storage and dual purpose cask models for residues and all kinds of radioactive wastes. The present paper proposes to illustrate how a company acting both as a cask designer and a carrier is key to the waste management issue and how it can support the waste management policy of nuclear waste producers through their operational choices. We will focus on the TN International technical solutions implemented to guarantee safe and secure transportation and storage solutions. We will describe different aspects of the cask design process, insisting on how it enables to fulfil both customer needs and regulation requirements. We will also mention the associated services developed by the AREVA Business Unit Logistics (TN International, TRANSNUCLEAR, MAINCO, and LMC) in order to manage transportation of liquid and solid waste towards interim or final storage sites. (authors)

  6. Biomethane as an Environment-Friendly Fuel for Municipal Transport

    NASA Astrophysics Data System (ADS)

    Papacz, Władysław J.

    2012-09-01

    An idea to replace the conventional engine fuel with biomethane has been described. The effects of such a swap have been analysed for municipal transport buses in Warsaw. The replacement has been shown to result in a significant reduction of toxic exhaust emissions and in economic savings. The analysis results have been presented in the form of tables and graphs.

  7. Thermal-hydraulic analysis of spent fuel storage systems

    SciTech Connect

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1987-01-01

    This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs.

  8. Transportation fuels from biomass via fast pyrolysis and hydroprocessing

    SciTech Connect

    Elliott, Douglas C.

    2013-09-21

    Biomass is a renewable source of carbon, which could provide a means to reduce the greenhouse gas impact from fossil fuels in the transportation sector. Biomass is the only renewable source of liquid fuels, which could displace petroleum-derived products. Fast pyrolysis is a method of direct thermochemical conversion (non-bioconversion) of biomass to a liquid product. Although the direct conversion product, called bio-oil, is liquid; it is not compatible with the fuel handling systems currently used for transportation. Upgrading the product via catalytic processing with hydrogen gas, hydroprocessing, is a means that has been demonstrated in the laboratory. By this processing the bio-oil can be deoxygenated to hydrocarbons, which can be useful replacements of the hydrocarbon distillates in petroleum. While the fast pyrolysis of biomass is presently commercial, the upgrading of the liquid product by hydroprocessing remains in development, although it is moving out of the laboratory into scaled-up process demonstration systems.

  9. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    SciTech Connect

    Wang, Jy-An John; Wang, Hong

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  10. 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.

  11. Multiphase transport in polymer electrolyte membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Gauthier, Eric D.

    Polymer electrolyte membrane fuel cells (PEMFCs) enable efficient conversion of fuels to electricity. They have enormous potential due to the high energy density of the fuels they utilize (hydrogen or alcohols). Power density is a major limitation to wide-scale introduction of PEMFCs. Power density in hydrogen fuel cells is limited by accumulation of water in what is termed fuel cell `flooding.' Flooding may occur in either the gas diffusion layer (GDL) or within the flow channels of the bipolar plate. These components comprise the electrodes of the fuel cell and balance transport of reactants/products with electrical conductivity. This thesis explores the role of electrode materials in the fuel cell and examines the fundamental connection between material properties and multiphase transport processes. Water is generated at the cathode catalyst layer. As liquid water accumulates it will utilize the largest pores in the GDL to go from the catalyst layer to the flow channels. Water collects to large pores via lateral transport at the interface between the GDL and catalyst layer. We have shown that water may be collected in these large pores from several centimeters away, suggesting that we could engineer the GDL to control flooding with careful placement and distribution of large flow-directing pores. Once liquid water is in the flow channels it forms slugs that block gas flow. The slugs are pushed along the channel by a pressure gradient that is dependent on the material wettability. The permeable nature of the GDL also plays a major role in slug growth and allowing bypass of gas between adjacent channels. Direct methanol fuel cells (DMFCs) have analogous multiphase flow issues where carbon dioxide bubbles accumulate, `blinding' regions of the fuel cell. This problem is fundamentally similar to water management in hydrogen fuel cells but with a gas/liquid phase inversion. Gas bubbles move laterally through the porous GDL and emerge to form large bubbles within the

  12. Understanding the transport processes in polymer electrolyte membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Cheah, May Jean

    Polymer electrolyte membrane (PEM) fuel cells are energy conversion devices suitable for automotive, stationary and portable applications. An engineering challenge that is hindering the widespread use of PEM fuel cells is the water management issue, where either a lack of water (resulting in membrane dehydration) or an excess accumulation of liquid water (resulting in fuel cell flooding) critically reduces the PEM fuel cell performance. The water management issue is addressed by this dissertation through the study of three transport processes occurring in PEM fuel cells. Water transport within the membrane is a combination of water diffusion down the water activity gradient and the dragging of water molecules by protons when there is a proton current, in a phenomenon termed electro-osmotic drag, EOD. The impact of water diffusion and EOD on the water flux across the membrane is reduced due to water transport resistance at the vapor/membrane interface. The redistribution of water inside the membrane by EOD causes an overall increase in the membrane resistance that regulates the current and thus EOD, thereby preventing membrane dehydration. Liquid water transport in the PEM fuel cell flow channel was examined at different gas flow regimes. At low gas Reynolds numbers, drops transitioned into slugs that are subsequently pushed out of the flow channel by the gas flow. The slug volume is dependent on the geometric shape, the surface wettability and the orientation (with respect to gravity) of the flow channel. The differential pressure required for slug motion primarily depends on the interfacial forces acting along the contact lines at the front and the back of the slug. At high gas Reynolds number, water is removed as a film or as drops depending on the flow channel surface wettability. The shape of growing drops at low and high Reynolds number can be described by a simple interfacial energy minimization model. Under flooding conditions, the fuel cell local current

  13. Transport Studies and Modeling in PEM Fuel Cells

    SciTech Connect

    Mittelsteadt, Cortney K.; Xu, Hui; Brawn, Shelly

    2014-07-30

    This project’s aim was to develop fuel cell components (i.e. membranes, gas-diffusion media (GDM), bipolar plates and flow fields) that possess specific properties (i.e. water transport and conductivity). A computational fluid dynamics model was developed to elucidate the effect of certain parameters on these specific properties. Ultimately, the model will be used to determine sensitivity of fuel cell performance to component properties to determine limiting components and to guide research. We have successfully reached our objectives and achieved most of the milestones of this project. We have designed and synthesized a variety of hydrocarbon block polymer membranes with lower equivalent weight, structure, chemistry, phase separation and process conditions. These membranes provide a broad selection with optimized water transport properties. We have also designed and constructed a variety of devices that are capable of accurately measuring the water transport properties (water uptake, water diffusivity and electro-osmatic drag) of these membranes. These transport properties are correlated to the membranes’ structures derived from X-ray and microscopy techniques to determine the structure-property relationship. We successfully integrated hydrocarbon membrane MEAs with a current distribution board (CBD) to study the impact of hydrocarbon membrane on water transport in fuel cells. We have designed and fabricated various GDM with varying substrate, diffusivity and micro-porous layers (MPL) and characterized their pore structure, tortuosity and hydrophobicity. We have derived a universal chart (MacMullin number as function of wet proofing and porosity) that can be used to characterize various GDM. The abovementioned GDMs have been evaluated in operating fuel cells; their performance is correlated to various pore structure, tortuosity and hydrophobicity of the GDM. Unfortunately, determining a universal relationship between the MacMullin number and these properties

  14. Implications of the Baltimore Rail Tunnel Fire for Full-Scale Testing of Shipping Casks

    SciTech Connect

    Halstead, R. J.; Dilger, F.

    2003-02-25

    The U.S. Nuclear Regulatory Commission (NRC) does not currently require full-scale physical testing of shipping casks as part of its certification process. Stakeholders have long urged NRC to require full-scale testing as part of certification. NRC is currently preparing a full-scale casktesting proposal as part of the Package Performance Study (PPS) that grew out of the NRC reexamination of the Modal Study. The State of Nevada and Clark County remain committed to the position that demonstration testing would not be an acceptable substitute for a combination of full-scale testing, scale-model tests, and computer simulation of each new cask design prior to certification. Based on previous analyses of cask testing issues, and on preliminary findings regarding the July 2001 Baltimore rail tunnel fire, the authors recommend that NRC prioritize extra-regulatory thermal testing of a large rail cask and the GA-4 truck cask under the PPS. The specific fire conditions and other aspects of the full-scale extra-regulatory tests recommended for the PPS are yet to be determined. NRC, in consultation with stakeholders, must consider past real-world accidents and computer simulations to establish temperature failure thresholds for cask containment and fuel cladding. The cost of extra-regulatory thermal testing is yet to be determined. The minimum cost for regulatory thermal testing of a legal-weight truck cask would likely be $3.3-3.8 million.

  15. Study of fuel systems for LH2-fueled subsonic transport aircraft, volume 1

    NASA Technical Reports Server (NTRS)

    Brewer, G. D.; Morris, R. E.; Davis, G. W.; Versaw, E. F.; Cunnington, G. R., Jr.; Riple, J. C.; Baerst, C. F.; Garmong, G.

    1978-01-01

    Several engine concepts examined to determine a preferred design which most effectively exploits the characteristics of hydrogen fuel in aircraft tanks received major emphasis. Many candidate designs of tank structure and cryogenic insulation systems were evaluated. Designs of all major elements of the aircraft fuel system including pumps, lines, valves, regulators, and heat exchangers received attention. Selected designs of boost pumps to be mounted in the LH2 tanks, and of a high pressure pump to be mounted on the engine were defined. A final design of LH2-fueled transport aircraft was established which incorporates a preferred design of fuel system. That aircraft was then compared with a conventionally fueled counterpart designed to equivalent technology standards.

  16. The resistance to impact of spent Magnox fuel transport flasks

    SciTech Connect

    Not Available

    1985-01-01

    This book completes the papers of the four-year programme of research and demonstrations embarked upon by the CEGB in 1981, culminating in the spectacular train crash at Old Dalby in July 1984. It explains the CEGB's operations in relation to the transportation of spent Magnox fuel. The public tests described in this book are more effective in improving public understanding and confidence than any amount of explanations could have been, raising the wider question of how best the scientific community can respond to the legitimate concerns of the man and woman in the street about the generating of electricity from nuclear power. The contents are: Taking care; irradiated fuel transport in the UK; programming for flask safety; the use of scale models in impact testing; flask analytical studies; drop test facilities; demonstration drop test; a study of flask transport impact hazards; impact of Magnox irradiated fuel transport flasks into rock and concrete; rail crash demonstration scenarios; horizontal impact testing of quarter scale flasks using masonry targets; horizontal crash testing and analysis of model flatrols; flatrol test; analysis of full scale impact into an abutment; analysis of primary impact forces in the train crash demonstration; horizontal impact tests of quarter scale Magnox flasks and stylised model locomotives; predictive estimates for behaviour in the train crash demonstration; design and organization of the crash; execution of the crash demonstration by British Rail; instrumentation for the train crash demonstration; photography for the crash demonstration; a summary of the CEGB's flask accident impact studies.

  17. A Study of Transport Airplane Crash-Resistant Fuel Systems

    NASA Technical Reports Server (NTRS)

    Jones, Lisa (Technical Monitor); Robertson, S. H.; Johnson, N. B.; Hall, D. S.; Rimson, I. J.

    2002-01-01

    This report presents the results of a study, funded by the Federal Aviation Administration (FAA), of transport airplane crash-resistant fuel system (CRFS). The report covers the historical studies related to aircraft crash fires and fuel containment concepts undertaken by the FAA, NASA, and the U.S. Army, which ultimately led to the current state of the art in CRFS technology. It describes the basic research, testing, field investigations and production efforts which have led to the highly successful military CRFS, which has saved many lives and reduced costs of accidents. Current CRFS technology used in transport category airplanes is defined and compared to the available state-of-the-art technology. The report provides information to the FAA and other government organizations which can help them plan their efforts to improve the state of crash fire protection in the transport airplane fleet. The report provides guidance to designers looking for information about CRFS design problems, analysis tools to use for product improvement, and a summary of current and proposed regulations for transport category airplane fuel systems.

  18. Fuel conservative propulsion concepts for future air transports

    NASA Technical Reports Server (NTRS)

    Gray, D. E.; Witherspoon, J. W.

    1976-01-01

    The results of a feasibility study of proposed fuel conservative propulsion concepts for air transports with an assumed Mach 0.8 cruise capability are summarized. All engines considered are based on projected 1985 technology. Operating fuel requirements, propulsion operating costs, and noise characteristics are compared with those of a present technology turbofan engine. The study indicates that an advanced Brayton cycle gas generator in a turbofan engine or geared to an advanced multibladed, small diameter propeller with a projected efficiency of 80% at Mach 0.8 offers the greatest potential for energy conservation.

  19. Study of LH2 fueled subsonic passenger transport aircraft

    NASA Technical Reports Server (NTRS)

    Brewer, G. D.; Morris, R. E.

    1976-01-01

    The potential of using liquid hydrogen as fuel in subsonic transport aircraft was investigated to explore an expanded matrix of passenger aircraft sizes. Aircraft capable of carrying 130 passengers 2,780 km (1500 n.mi.); 200 passengers 5,560 km (3000 n.mi.); and 400 passengers on a 9,265 km (5000 n.mi.) radius mission, were designed parametrically. Both liquid hydrogen and conventionally fueled versions were generated for each payload/range in order that comparisons could be made. Aircraft in each mission category were compared on the basis of weight, size, cost, energy utilization, and noise.

  20. Steady natural convection heat transfer experiments in a horizontal annulus for the United States Spent Fuel Shipping Cask Technology Program. [LMFBR

    SciTech Connect

    Boyd, R. D.

    1981-04-01

    This experimental study deals with the measurement of the heat transfer across a horizontal annulus which is formed by an inner hexagonal cylinder and an outer concentric circular cylinder. The geometry simulates, in two dimensions, a liquid metal fast breeder reactor radioactive fuel subassembly inside a shipping container. This geometry is also similar to a radioactive fuel pin inside a horizontal reactor subassembly. The objective of the experiments is to measure the local and mean heat transfer at the surface of the inner hexagonal cylinder.

  1. Fuel Cell System for Transportation -- 2005 Cost Estimate

    SciTech Connect

    Wheeler, D.

    2006-10-01

    Independent review report of the methodology used by TIAX to estimate the cost of producing PEM fuel cells using 2005 cell stack technology. The U.S. Department of Energy (DOE) Hydrogen, Fuel Cells and Infrastructure Technologies Program Manager asked the National Renewable Energy Laboratory (NREL) to commission an independent review of the 2005 TIAX cost analysis for fuel cell production. The NREL Systems Integrator is responsible for conducting independent reviews of progress toward meeting the DOE Hydrogen Program (the Program) technical targets. An important technical target of the Program is the proton exchange membrane (PEM) fuel cell cost in terms of dollars per kilowatt ($/kW). The Program's Multi-Year Program Research, Development, and Demonstration Plan established $125/kW as the 2005 technical target. Over the last several years, the Program has contracted with TIAX, LLC (TIAX) to produce estimates of the high volume cost of PEM fuel cell production for transportation use. Since no manufacturer is yet producing PEM fuel cells in the quantities needed for an initial hydrogen-based transportation economy, these estimates are necessary for DOE to gauge progress toward meeting its targets. For a PEM fuel cell system configuration developed by Argonne National Laboratory, TIAX estimated the total cost to be $108/kW, based on assumptions of 500,000 units per year produced with 2005 cell stack technology, vertical integration of cell stack manufacturing, and balance-of-plant (BOP) components purchased from a supplier network. Furthermore, TIAX conducted a Monte Carlo analysis by varying ten key parameters over a wide range of values and estimated with 98% certainty that the mean PEM fuel cell system cost would be below DOE's 2005 target of $125/kW. NREL commissioned DJW TECHNOLOGY, LLC to form an Independent Review Team (the Team) of industry fuel cell experts and to evaluate the cost estimation process and the results reported by TIAX. The results of this

  2. Issues in International Energy Consumption Analysis: Chinese Transportation Fuel Demand

    EIA Publications

    2014-01-01

    Since the 1990s, China has experienced tremendous growth in its transportation sector. By the end of 2010, China's road infrastructure had emerged as the second-largest transportation system in the world after the United States. Passenger vehicle sales are dramatically increasing from a little more than half a million in 2000, to 3.7 million in 2005, to 13.8 million in 2010. This represents a twenty-fold increase from 2000 to 2010. The unprecedented motorization development in China led to a significant increase in oil demand, which requires China to import progressively more petroleum from other countries, with its share of petroleum imports exceeding 50% of total petroleum demand since 2009. In response to growing oil import dependency, the Chinese government is adopting a broad range of policies, including promotion of fuel-efficient vehicles, fuel conservation, increasing investments in oil resources around the world, and many others.

  3. National Program Plan Fuel Cells in Transportation. Executive Summary

    SciTech Connect

    Not Available

    1993-02-01

    Fuel cells are being developed for application in the transportation sector because they will convert hydrogen to electric power at high efficiencies with virtually no detrimental environmental impact. To realize these energy, environmental, and economic benefits, developers of FCVs need to (1) reduce the size and weight of current designs, (2) develop fuel cell propulsion systems with rapid start-up and greater load-following capability, (3) reduce system cost and/or improve performance, and (4) utilize alternative fuels to a large extent. This Plan addresses the FCV-related requirements of the Energy Act, describing a development program for light- and heavy-duty propulsion systems, a basic R&D program on fuel cell technology that is separate from, but feeds into, the system development activities, and supporting analyses. Implementation of the Program Plan by means of industry/government alliances will accelerate the commercialization of FCVs. In the long term, the successful deployment of large numbers of FCVs promises to eliminate the transportation sector as a major contributor to the nation`s environmental problems.

  4. Salt transport extraction of transuranium elements from lwr fuel

    DOEpatents

    Pierce, R. Dean; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

  5. Salt transport extraction of transuranium elements from LWR fuel

    DOEpatents

    Pierce, R.D.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Miller, W.E.

    1992-11-03

    A process is described for separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl[sub 2] and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750 C to about 850 C to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl[sub 2] having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO[sub 2]. The Ca metal and CaCl[sub 2] is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including MgCl[sub 2] to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy. 2 figs.

  6. Management of the spent fuel elements of the thorium high temperature reactor THTR-300

    SciTech Connect

    Quaassdorff, P.; Mielisch, M.; Dietrich, G.; Heske, M.; Jacobsen, W.

    1995-12-31

    In a world-wide unique campaign ca. 620,000 spent fuel elements of the thorium high temperature reactor THTR 300 which is being decommissioned, were being transferred within a short period of time to the Ahaus fuel element interim store (BZA) for interim storage. In order to optimize the technical and logistic procedures as part of the pre-decommissioning operation in 1992 and 1993, 42,000 fuel elements which had already been removed from the reactor core were transferred to Ahaus in transport and storage casks of the CASTOR THTR/AVR type that have been specially designed for this purpose. The experiences gained with loading, processing and transport of 20 transport and storage casks during this optimization and testing period led the team to expect a smooth management of the remaining fuel elements. In January 1994, the routine operation of the outward transfer commenced. Until mid-November 1994, 554,400 spent fuel elements were transferred outward into altogether 264 transport and storage casks of the CASTOR THTR/AVR type and transported to Ahaus for interim storage. This was followed by processing of another 21 transport and storage casks until April 1995, accommodating damaged fuel elements and special elements. The work mentioned above was performed by SFEAG Kernenergie GmbH, Essen, on behalf of the reactor operator Hochtemperatur-Kernkraftwerk GmbH, Hamm. The removal of the nuclear fuel from the thorium high temperature reactor THTR-300 marks the completion of the first part of the necessary actions for the decommissioning of the reactor (safe enclosure).

  7. Software requirements definition Shipping Cask Analysis System (SCANS)

    SciTech Connect

    Johnson, G.L.; Serbin, R.

    1985-07-21

    The US Nuclear Regulatory Commission (NRC) staff reviews the technical adequacy of applications for certification of designs of shipping casks for spent nuclear fuel. In order to confirm an acceptable design, the NRC staff may perform independent calculations. The current NRC procedure for confirming cask design analyses is laborious and tedious. Most of the work is currently done by hand or through the use of a remote computer network. The time required to certify a cask can be long. The review process may vary somewhat with the engineer doing the reviewing. Similarly, the documentation on the results of the review can also vary with the reviewer. To increase the efficiency of this certification process, LLNL was requested to design and write an integrated set of user-oriented, interactive computer programs for a personal microcomputer. The system is known as the NRC Shipping Cask Analysis System (SCANS). The computer codes and the software system supporting these codes are being developed and maintained for the NRC by LLNL. The objective of this system is generally to lessen the time and effort needed to review an application. Additionally, an objective of the system is to assure standardized methods and documentation of the confirmatory analyses used in the review of these cask designs. A software system should be designed based on NRC-defined requirements contained in a requirements document. The requirements document is a statement of a project's wants and needs as the users and implementers jointly understand them. The requirements document states the desired end products (i.e. WHAT's) of the project, not HOW the project provides them. This document describes the wants and needs for the SCANS system. 1 fig., 3 tabs.

  8. Results of the first thirty foot drop test of the MOSAIK KfK cask

    SciTech Connect

    Sorenson, K.B.; Salzbrenner, R.; Wellman, G.; Uncapher, W.; Bobbe, J.

    1991-01-01

    The MOSAIK KfK cask, a ductile cast iron (DCI) nuclear material transportation cask donated to Sandia by Gesellschaft fur NuklearService (GNS), was drop tested on June 25, 1990 in Albuquerque, New Mexico. Conditions of the test were; a 30 ft. drop without impact limiters onto an unyielding target, cask metal temperature {minus}16F or below, and a 0.75 inch deep flaw machined into the cask wall at the location of the highest tensile stress. The drop test was successful as judged by inspection of the machined flaw which showed no crack initiation. This drop test, in the first in a series, was designed to demonstrate the viability of using a fracture mechanics approach to design cask fabricated from ferritic materials. In addition, the test demonstrated that a DCI cask can withstand severe impacts under accident-type conditions without failing in a brittle mode. The drop test parameters were designed to produce high decelerations and yield-level stresses in the cask wall. The measured rigid body deceleration was approximately 800 gs. This compares with decelerations of 100 to 300 gs for drop tests of casks with impact limiters. The time to peak load was 1.2 to 2.8 msec., compared to 20 to 40 msec for casks dropped with impact limiters. The maximum strain during the drop test was 1400 microstrain, which equates to a maximum tensile stress of about 37000 psi. This level of stress slightly exceeds the static yield strength and is about 80% of the dynamic yield strength. The test results of this initial drop test are discussed in detail in this paper.

  9. 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.

  10. Understanding of ammonia transport in PEM fuel cells

    NASA Astrophysics Data System (ADS)

    Jung, Myunghee

    This dissertation investigates ammonia (NH3) as a fuel contaminant to the anode in Proton Exchange Membrane Fuel Cells (PEMFCs). Since NH 3 is fed to the anode in a gas phase and transferred to the cathode, the effect of a contaminant is distributed through MEA and quite complicated. This study is focused on the investigation of mechanism of NH3 transport and the isolation of multiple effects to degrade the performance of fuel cell. An External Reference Electrode (ERE) was employed to decouple the effect of individual electrode and explain the mechanism of NH3 contamination. A mechanism of NH3 transport is proposed and supported by data for various inlet conditions in a N2/N2 laboratory-scale fuel cell at Open Circuit Conditions (OCC). With a commercialized GORE(TM) PRIMEA RTM 5631 MEAs at 70°C, data were obtained utilizing a material balance technique, which uses an ion selective electrode (ISE) to determine the concentration of ammonium ion in the process streams. The results indicate that ammonia is not transported across the membrane when the feeds to both electrodes are dry. However, with humidified feeds ammonia was transported from the anode to the cathode. The data also indicate the water content of in the MEA is the critical factor that causes NH3 crossover in the MEA. Diffusion coefficients of NH3 in MEA are also calculated at different relative humilities. An ERE was developed for PEM fuel cell by using a NafionRTM strip which was used to understand contamination mechanism. The voltage of anode electrode relative to ERE was measured during a polarization curve. The data showed the measurement of individual electrode potential was extremely affected by the misalignment between two electrodes. We compare the overpotential measured from the reference electrode and the calculated overpotential from subtracting the cell voltages between neat hydrogen and a 25 ppm CO in H 2 stream at same current. The studies indicated that the overpotentials obtained from

  11. Advanced fuel cells for transportation applications. Final report

    SciTech Connect

    1998-02-10

    This Research and Development (R and D) contract was directed at developing an advanced technology compressor/expander for supplying compressed air to Proton Exchange Membrane (PEM) fuel cells in transportation applications. The objective of this project was to develop a low-cost high-efficiency long-life lubrication-free integrated compressor/expander utilizing scroll technology. The goal of this compressor/expander was to be capable of providing compressed air over the flow and pressure ranges required for the operation of 50 kW PEM fuel cells in transportation applications. The desired ranges of flow, pressure, and other performance parameters were outlined in a set of guidelines provided by DOE. The project consisted of the design, fabrication, and test of a prototype compressor/expander module. The scroll CEM development program summarized in this report has been very successful, demonstrating that scroll technology is a leading candidate for automotive fuel cell compressor/expanders. The objectives of the program are: develop an integrated scroll CEM; demonstrate efficiency and capacity goals; demonstrate manufacturability and cost goals; and evaluate operating envelope. In summary, while the scroll CEM program did not demonstrate a level of performance as high as the DOE guidelines in all cases, it did meet the overriding objectives of the program. A fully-integrated, low-cost CEM was developed that demonstrated high efficiency and reliable operation throughout the test program. 26 figs., 13 tabs.

  12. Direct-hydrogen-fueled proton-exchange-membrane fuel cell system for transportation applications

    SciTech Connect

    Oei, D.; Adams, J.A.; Kinnelly, A.A.

    1997-07-01

    In partial fulfillment of the U.S. Department of Energy Contract No. DE-ACO2-94CE50389, {open_quotes}Direct Hydrogen-Fueled Proton-Exchange-Membrane (PEM) Fuel Cell System for Transportation Applications{close_quotes}, this conceptual vehicle design report addresses the design and packaging of battery augmented fuel cell powertrain vehicles. This report supplements the {open_quotes}Conceptual Vehicle Design Report - Pure Fuel Cell Powertrain Vehicle{close_quotes} and includes a cost study of the fuel cell power system. The three classes of vehicles considered in this design and packaging exercise are the same vehicle classes that were studied in the previous report: the Aspire, representing the small vehicle class; the AIV (Aluminum Intensive Vehicle) Sable, representing the mid-size vehicle; and the E-150 Econoline, representing the van-size class. A preliminary PEM fuel cell power system manufacturing cost study is also presented. As in the case of the previous report concerning the {open_quotes}Pure Fuel Cell Powertrain Vehicle{close_quotes}, the same assumptions are made for the fuel cell power system. These assumptions are fuel cell system power densities of 0.33 kW/ka and 0.33 kW/l, platinum catalyst loading of less than or equal to 0.25 mg/cm{sup 2} total, and hydrogen tanks containing compressed gaseous hydrogen under 340 atm (5000 psia) pressure. The batteries considered for power augmentation of the fuel cell vehicle are based on the Ford Hybrid Electric Vehicle (HEV) program. These are state-of-the-art high power lead acid batteries with power densities ranging from 0.8 kW/kg to 2 kW/kg. The results reported here show that battery augmentation provides the fuel cell vehicle with a power source to meet instant high power demand for acceleration and start-up. Based on the assumptions made in this report, the packaging of the battery augmented fuel cell vehicle appears to be as feasible as the packaging of the pure fuel cell powered vehicle.

  13. Combustion chemistry and an evolving transportation fuel environment.

    SciTech Connect

    Taatjes, Craig A.

    2010-05-01

    The world currently faces tremendous energy challenges stemming from the need to curb potentially catastrophic anthropogenic climate change. In addition, many nations, including the United States, recognize increasing political and economic risks associated with dependence on uncertain and limited energy sources. For these and other reasons the chemical composition of transportation fuels is changing, both through introduction of nontraditional fossil sources, such as oil sands-derived fuels in the US stream, and through broader exploration of biofuels. At the same time the need for clean and efficient combustion is leading engine research towards advanced low-temperature combustion strategies that are increasingly sensitive to this changing fuel chemistry, particularly in the areas of pollutant formation and autoignition. I will highlight the new demands that advanced engine technologies and evolving fuel composition place on investigations of fundamental reaction chemistry. I will focus on recent progress in measuring product formation in elementary reactions by tunable synchrotron photoionization, on the elucidation of pressure-dependent effects in the reactions of alkyl and substituted alkyl radicals with O{sub 2}, and on new combined efforts in fundamental combustion chemistry and engine performance studies of novel potential biofuels.

  14. Evaluation of Simplified Models for Estimating Public Dose from Spent Nuclear Fuel Shipments

    SciTech Connect

    Connolly, Kevin J.; Radulescu, Georgeta

    2015-01-01

    This paper investigates the dose rate as a function of distance from a representative high-capacity SNF rail-type transportation cask. It uses the SCALE suite of radiation transport modeling and simulation codes to determine neutron and gamma radiation dose rates. The SCALE calculated dose rate is compared with simplified analytical methods historically used for these calculations. The SCALE dose rate calculation presented in this paper employs a very detailed transportation cask model (e.g., pin-by-pin modeling of fuel assembly) and a new hybrid computational transport method. Because it includes pin-level heterogeneity and models ample air and soil outside the cask to simulate scattering of gamma and neutron radiation, this detailed SCALE model is expected to yield more accurate results than previously used models which made more simplistic assumptions (e.g., fuel assembly treated as a point or line source, simple 1-D model of environment outside of cask). The results in this paper are preliminary and, as progress is made on developing and validating improved models, results may be subject to change as models and estimates become more refined and better information leads to more accurate assumptions.

  15. Cask system maintenance in the Federal Waste Management System

    SciTech Connect

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

    1991-01-01

    In early 1988, in support of the development of the transportation system for the Office of Civilian Radioactive Waste Management System (OCRWM), a feasibility study was undertaken to define a the concept for a stand-alone, green-field'' facility for maintaining the Federal Waste Management System (FWMS) casks. This study provided and initial layout facility design, an estimate of the construction costs, and an acquisition schedule for a Cask Maintenance Facility (CMF). It also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrievable Storage (MRS) facility. The data, design, and estimated costs derived from the study have been organized for use in the total transportation system decision-making process. Most importantly, they also provide a foundation for continuing design and planning efforts. The feasibility study was based on an assumed stand-alone, green-field'' configuration. This design approach provides a comprehensive design evaluation, to guide the development of a cost estimate and to permit flexibility in locating the facility. The following sections provide background information on cask system maintenance, briefly summarizes some of the functional requirements that a CMF must satisfy, provides a physical description of the CMF, briefly discusses the cost and schedule estimates and then reviews the findings of the efforts undertaken since the feasibility study was completed. 15 refs., 3 figs.

  16. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    SciTech Connect

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  17. Stress analysis of closure bolts for shipping casks

    SciTech Connect

    Mok, G.C.; Fischer, L.E. ); Hsu, S.T. )

    1993-01-01

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

  18. 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...

  19. Vestibule and Cask Preparation Mechanical Handling Calculation

    SciTech Connect

    N. Ambre

    2004-05-26

    The scope of this document is to develop the size, operational envelopes, and major requirements of the equipment to be used in the vestibule, cask preparation area, and the crane maintenance area of the Fuel Handling Facility. This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAIC Company L.L.C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC28-01R W12101'' (Ref. 167124). This correspondence was appended by further correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC28-01R W12101; TDL No. 04-024'' (Ref. 16875 1). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process.

  20. Fischer-Tropsch slurry catalysts for selective transportation fuel production

    SciTech Connect

    Carroll, W.E.; Cilen, N.; Withers, H.P. Jr.

    1986-01-01

    The future use of coal as a source of conventional transportation fuel will depend on the development of an economical and energy efficient liquefaction process. Technologies that have been commercially proven or that are close to commercialization include the fixed- and fluidized-bed Fischer-Tropsch (FT) synthesis, methanol synthesis (fixed-bed and slurry-phase) and the Mobil methanol-to-gasoline process. Of these technologies, the Fischer-Tropsch hydrocarbon synthesis produces the widest slate of products and has been in operation for the longest period.

  1. The role of technology as air transportation faces the fuel situation

    NASA Technical Reports Server (NTRS)

    Driver, C.

    1980-01-01

    Perspectives on the air transportation fuel stituation are discussed including intercity air traffic, airline fuel consumption, fuel price effects on ticket price, and projected traffic and fuel useage between now and the year 2000. Actions taken by the airlines to reduce consumption are reviewed, as well as efforts currently underway to improve fuel consumption. Longer range technology payoffs resulting from NASA research programs are reviewed and results from studies on the use of alternate fuels are discussed.

  2. Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

    SciTech Connect

    Marshall, William BJ J; Wagner, John C

    2012-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

  3. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    SciTech Connect

    J. Bisset

    2005-02-14

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known.

  4. Energy analysis and break-even distance for transportation for biofuels in comparison to fossil fuels

    Technology Transfer Automated Retrieval System (TEKTRAN)

    In the present analysis various forms fuel from biomass and fossil sources, their mass and energy densities, and their break-even transportation distances to transport them effectively were analyzed. This study gives an insight on how many times more energy spent on transporting the fuels to differe...

  5. Capability of environmental sampling to detect undeclared cask openings

    SciTech Connect

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

    1997-12-01

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

  6. HEAT AND WATER TRANSPORT IN A POLYMER ELECTROLYTE FUEL CELL

    SciTech Connect

    Mukherjee, Partha P

    2010-01-01

    In the present scenario of a global initiative toward a sustainable energy future, the polymer electrolyte fuel cell (PEFC) has emerged as one of the most promising alternative energy conversion devices for various applications. Despite tremendous progress in recent years, a pivotal performance limitation in the PEFC comes from liquid water transport and the resulting flooding phenomena. Liquid water blocks the open pore space in the electrode and the fibrous diffusion layer leading to hindered oxygen transport. The electrode is also the only component in the entire PEFC sandwich which produces waste heat from the electrochemical reaction. The cathode electrode, being the host to several competing transport mechanisms, plays a crucial role in the overall PEFC performance limitation. In this work, an electrode model is presented in order to elucidate the coupled heat and water transport mechanisms. Two scenarios are specifically considered: (1) conventional, Nafion impregnated, three-phase electrode with the hydrated polymeric membrane phase as the conveyer of protons where local electro-neutrality prevails; and (2) ultra-thin, two-phase, nano-structured electrode without the presence of ionomeric phase where charge accumulation due to electro-statics in the vicinity of the membrane-CL interface becomes important. The electrode model includes a physical description of heat and water balance along with electrochemical performance analysis in order to study the influence of electro-statics/electro-migration and phase change on the PEFC electrode performance.

  7. 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.

  8. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect

    Not Available

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  9. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect

    Not Available

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  10. Spent Fuel and High-Level Radioactive Waste Transportation Report

    SciTech Connect

    Not Available

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  11. DOE/NEAMS AMP CAMP I 2010 - multi species transport in metal fuels

    SciTech Connect

    Dilts, Gary A

    2011-01-21

    Essential aspects from the literature of metal nuclear fuel alloys and modeling the transport of constituents therein are discussed. The essential mathematical problem is described along with relevant issues for implementation of solution algorithms in the AMP nuclear fuel code.

  12. Basic Research Needs for Clean and Efficient Combustion of 21st Century Transportation Fuels

    SciTech Connect

    McIlroy, A.; McRae, G.; Sick, V.; Siebers, D. L.; Westbrook, C. K.; Smith, P. J.; Taatjes, C.; Trouve, A.; Wagner, A. F.; Rohlfing, E.; Manley, D.; Tully, F.; Hilderbrandt, R.; Green, W.; Marceau, D.; O'Neal, J.; Lyday, M.; Cebulski, F.; Garcia, T. R.; Strong, D.

    2006-11-01

    To identify basic research needs and opportunities underlying utilization of evolving transportation fuels, with a focus on new or emerging science challenges that have the potential for significant long-term impact on fuel efficiency and emissions.

  13. Aluminum-fueled rockets for the space transportation system

    NASA Technical Reports Server (NTRS)

    Cutler, Andrew Hall

    1992-01-01

    Aluminum-fueled engines, used to propel orbital transfer vehicles (OTV's), offer benefits to the Space Transportation System (STS) if scrap aluminum can be scavenged at a reasonable cost. Aluminum scavenged from Space Shuttle external tanks could replace propellants hauled from Earth, thus allowing more payloads to be sent to their final destinations at the same Shuttle launch rate. To allow OTV use of aluminum fuel, two new items would be required: a facility to reprocess aluminum from external tanks and an engine for the OTV which could burn aluminum. Design of the orbital transfer vehicle would have to differ substantially from current concepts for it to carry and use the aluminum fuel. The aluminum reprocessing facility would probably have a mass of under 15 metric tons and would probably cost less that $200,000,000. Development of an aluminum-burning engine would no doubt be extremely expensive (1 to 2 billion dollars), but this amount would be adequately repaid by increased STS throughput. Engine production cost is difficult to estimate, but even an extremely high cost (e.g., $250,000,000 per engine) would not significantly increase orbit-raising expenses.

  14. SCANS (Shipping Cask ANalysis System) a microcomputer-based analysis system for shipping cask design review: User`s manual to Version 3a. Volume 1, Revision 2

    SciTech Connect

    Mok, G.C.; Thomas, G.R.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L.

    1998-03-01

    SCANS (Shipping Cask ANalysis System) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent fuel shipping casks. SCANS is an easy-to-use system that calculates the global response to impact loads, pressure loads and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. SCANS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests. Analysis options are based on regulatory cases described in the Code of Federal Regulations 10 CFR 71 and Regulatory Guides published by the US Nuclear Regulatory Commission in 1977 and 1978.

  15. Operational considerations in specifying legal weight vehicles for the highway transport of spent nuclear fuel

    SciTech Connect

    Hill, C V; Rutenkroger, E O; Ratledge, J E

    1990-01-01

    This paper presents the results of a research project in which tractor manufacturers and carrier companies were interviewed to gather information on operational concerns in specifying a tractor to haul legal weight spent fuel casks. The system was assumed to operate very close to the 80,000 pound legal weight limit. Safety, performance, reliability, and maintainability of equipment were factors given particular attention. The interaction between driver fatigue, safety, and equipment was also discussed. Innovative operating strategies that could save weight were discussed. The paper concluded that operational considerations require that planners working with standard off-the-shelf tractor equipment should allow at least 17,350 pounds for the weight of the tractor as a starting point from which further weight reduction analysis can proceed. 4 refs., 3 figs.

  16. Solid oxide fuel cells for transportation: A clean, efficient alternative for propulsion

    NASA Astrophysics Data System (ADS)

    Kumar, R.; Krumpelt, M.; Myles, K. M.

    Fuel cells show great promise for providing clean and efficient transportation power. Of the fuel cell propulsion systems under investigation, the solid oxide fuel cell (SOFC) is particularly attractive for heavy duty transportation applications that have a relatively long duty cycle, such as locomotives, trucks, and barges. Advantages of the SOFC include a simple, compact system configuration; inherent fuel flexibility for hydrocarbon and alternative fuels; and minimal water management. The specific advantages of the SOFC for powering a railroad locomotive are examined. Feasibility, practicality, and safety concerns regarding SOFCs in transportation applications are discussed.

  17. Shielding Analysis of the 5320 Shipping Cask

    SciTech Connect

    Blanchard, A.; Nathan, S.

    1998-05-01

    The purpose of this work is to demonstrate that the 5320 shipping cask meets Federal regulations for maximum radiation dose rates when loaded with the intended plutonium oxide cargo. It should be emphasized that the 5320 is an existing cask, and therefore this work represents confirmatory analysis rather than design analysis.

  18. Performance of bolted closure joint elastomers under cask aging conditions

    SciTech Connect

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

    2015-07-23

    The bolted closure joint of a bare spent fuel cask is susceptible to age-related degradation and potential loss of confinement function under long-term storage conditions. Elastomeric seals, a component of the joint typically used to facilitate leak testing of the primary seal that includes the metallic seal and bolting, is susceptible to degradation over time by several mechanisms, principally via thermo-oxidation, stress-relaxation, and radiolytic degradation under time and temperature condition. Irradiation and thermal exposure testing and evaluation of an ethylene-propylene diene monomer (EPDM) elastomeric seal material similar to that used in the CASTOR® V/21 cask for a matrix of temperature and radiation exposure conditions relevant to the cask extended storage conditions, and development of semiempirical predictive models for loss of sealing force is in progress. A special insert was developed to allow Compressive Stress Relaxation (CSR) measurements before and after the irradiation and/or thermal exposure without unloading the elastomer. A condition of the loss of sealing force for the onset of leakage was suggested. The experimentation and modeling being performed could enable acquisition of extensive coupled aging data as well as an estimation of the timeframe when loss of sealing function under aging (temperature/radiation) conditions may occur.

  19. USED FUEL RAIL SHOCK AND VIBRATION TESTING OPTIONS ANALYSIS

    SciTech Connect

    Ross, Steven B.; Best, Ralph E.; Klymyshyn, Nicholas A.; Jensen, Philip J.; Maheras, Steven J.

    2014-09-29

    The objective of the rail shock and vibration tests is to complete the framework needed to quantify loads of fuel assembly components that are necessary to guide materials research and establish a technical basis for review organizations such as the U.S. Nuclear Regulatory Commission (NRC). A significant body of experimental and numerical modeling data exists to quantify loads and failure limits applicable to normal conditions of transport (NCT) rail transport, but the data are based on assumptions that can only be verified through experimental testing. The test options presented in this report represent possible paths for acquiring the data that are needed to confirm the assumptions of previous work, validate modeling methods that will be needed for evaluating transported fuel on a case-by-case basis, and inform material test campaigns on the anticipated range of fuel loading. The ultimate goal of this testing is to close all of the existing knowledge gaps related to the loading of used fuel under NCT conditions and inform the experiments and analysis program on specific endpoints for their research. The options include tests that would use an actual railcar, surrogate assemblies, and real or simulated rail transportation casks. The railcar carrying the cradle, cask, and surrogate fuel assembly payload would be moved in a train operating over rail track modified or selected to impart shock and vibration forces that occur during normal rail transportation. Computer modeling would be used to help design surrogates that may be needed for a rail cask, a cask’s internal basket, and a transport cradle. The objective of the design of surrogate components would be to provide a test platform that effectively simulates responses to rail shock and vibration loads that would be exhibited by state-of-the-art rail cask, basket, and/or cradle structures. The computer models would also be used to help determine the placement of instrumentation (accelerometers and strain gauges

  20. NREL - Advanced Vehicles and Fuels Basics - Center for Transportation Technologies and Systems 2010

    SciTech Connect

    2010-01-01

    We can improve the fuel economy of our cars, trucks, and buses by designing them to use the energy in fuels more efficiently. Researchers at the National Renewable Energy Laboratory (NREL) are helping the nation achieve these goals by developing transportation technologies like: advanced vehicle systems and components; alternative fuels; as well as fuel cells, hybrid electric, and plug-in hybrid vehicles. For a text version of this video visit http://www.nrel.gov/learning/advanced_vehicles_fuels.html

  1. NREL - Advanced Vehicles and Fuels Basics - Center for Transportation Technologies and Systems 2010

    ScienceCinema

    None

    2013-05-29

    We can improve the fuel economy of our cars, trucks, and buses by designing them to use the energy in fuels more efficiently. Researchers at the National Renewable Energy Laboratory (NREL) are helping the nation achieve these goals by developing transportation technologies like: advanced vehicle systems and components; alternative fuels; as well as fuel cells, hybrid electric, and plug-in hybrid vehicles. For a text version of this video visit http://www.nrel.gov/learning/advanced_vehicles_fuels.html

  2. Packaging design criteria for the MCO cask

    SciTech Connect

    Edwards, W.S.

    1996-09-11

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the K Basins to a Canister Storage Building in the 200 East Area. The purpose of this packaging design criteria is to provide criteria for the design,fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multiple Canister Overpacks.

  3. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    SciTech Connect

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  4. REDUCING ULTRA-CLEAN TRANSPORTATION FUEL COSTS WITH HYMELT HYDROGEN

    SciTech Connect

    Donald P. Malone; William R. Renner

    2005-07-01

    Phase I of the work to be done under this agreement consisted of conducting atmospheric gasification of coal using the HyMelt technology to produce separate hydrogen rich and carbon monoxide rich product streams. In addition smaller quantities of petroleum coke and a low value refinery stream were gasified. Phase II of the work to be done under this agreement, consists of gasification of the above-mentioned feeds at a gasifier pressure of approximately 5 bar. The results of this work will be used to evaluate the technical and economic aspects of producing ultra-clean transportation fuels using the HyMelt technology in existing and proposed refinery configurations. This report describes activities for the ninth quarter of work performed under this agreement. The design of the vessel for pressure testing has been completed. The design will be finalized and purchased in the next quarter.

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

    SciTech Connect

    2014-02-27

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

  6. Packaging Design Criteria for the MCO Cask

    SciTech Connect

    FLANAGAN, B.D.

    2000-08-01

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

  7. Transient analysis of water transport in PEM fuel cells

    NASA Astrophysics Data System (ADS)

    Yan, Wei-Mon; Chu, Hsin-Sen; Chen, Jian-Yao; Soong, Chyi-Yeou; Chen, Falin

    This paper theoretically studies the water transport phenomena in PEM fuel cells, mainly investigating the transient behavior in the gas diffusion layer (GDL), catalyst layer (CL) and proton exchange membrane (PEM). In the PEM, both diffusion and electro-osmosis processes are considered, while in the GDL and CL, only diffusion process is taken into account. The process of water uptake is employed to account for the water transport at the interface between the PEM and CL. The results indicate that the water content in the PEM and the time for reaching the steady state in the start-up process are influenced by the humidification constant, k, the humidification, and the thickness of PEM. The rise of the k increases the water content in the membrane and shortens the time for reaching the steady state. Insufficient humidification causes relatively small water content and long steady time. When the PEM is thinner, the water is more uniformly distributed, the water content gets higher, and the time for reaching the steady state is distinctly shorter.

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

    SciTech Connect

    1995-03-01

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

  9. Conclusions and recommendations. [for problems in energy situation, air transportation, and hydrogen fuel

    NASA Technical Reports Server (NTRS)

    1973-01-01

    Conclusions and recommendations are presented for an analysis of the total energy situation; the effect of the energy problem on air transportation; and hydrogen fuel for aircraft. Properties and production costs of fuels, future prediction for energy and transportation, and economic aspects of hydrogen production are appended.

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

    SciTech Connect

    Edwards, W.S.

    1997-07-14

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

  11. 14 CFR 129.23 - Transport category cargo service airplanes: Increased zero fuel and landing weights.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... fuel and landing weights. (a) Notwithstanding the applicable structural provisions of the transport... fuel weight (maximum weight of the airplane with no disposable fuel and oil) and the structural landing... Administrator finds that— (1) The increase is not likely to reduce seriously the structural strength; (2)...

  12. 14 CFR 129.23 - Transport category cargo service airplanes: Increased zero fuel and landing weights.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... fuel and landing weights. (a) Notwithstanding the applicable structural provisions of the transport... fuel weight (maximum weight of the airplane with no disposable fuel and oil) and the structural landing... Administrator finds that— (1) The increase is not likely to reduce seriously the structural strength; (2)...

  13. 14 CFR 129.23 - Transport category cargo service airplanes: Increased zero fuel and landing weights.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... fuel and landing weights. (a) Notwithstanding the applicable structural provisions of the transport... fuel weight (maximum weight of the airplane with no disposable fuel and oil) and the structural landing... Administrator finds that— (1) The increase is not likely to reduce seriously the structural strength; (2)...

  14. 14 CFR 129.23 - Transport category cargo service airplanes: Increased zero fuel and landing weights.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... fuel and landing weights. (a) Notwithstanding the applicable structural provisions of the transport... fuel weight (maximum weight of the airplane with no disposable fuel and oil) and the structural landing... Administrator finds that— (1) The increase is not likely to reduce seriously the structural strength; (2)...

  15. Near-Site Transportation Infrastructure Project

    SciTech Connect

    Viebrock, J.M.; Mote, N. )

    1992-02-01

    There are 122 commercial nuclear facilities from which spent nuclear fuel will be accepted by the Federal Waste Management System (FWMS). Since some facilities share common sites and some facilities are on adjacent sites, 76 sites were identified for the Near-Site Transportation Infrastructure (NSTI) project. The objective of the NSTI project was to identify the options available for transportation of spent-fuel casks from each of these commercial nuclear facility sites to the main transportation routes -- interstate highways, commercial rail lines and navigable waterways available for commercial use. The near-site transportation infrastructure from each site was assessed, based on observation of technical features identified during a survey of the routes and facilities plus data collected from referenced information sources. The potential for refurbishment of transportation facilities which are not currently operational was also assessed, as was the potential for establishing new transportation facilities.

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

    SciTech Connect

    Guskov, Vladimir; Korotkov, Gennady; Barnes, Ella; Snipes, Randy

    2007-07-01

    Available in abstract form only. Full text of publication follows: In recent ten years a new technology of management of irradiated nuclear fuel (SNF) at the final stage of fuel cycle has been intensely developing on a basis of a new type of casks used for interim storage of SNF and subsequent transportation therein to the place of processing, further storage or final disposal. This technology stems from the concept of a protective cask which provides preservation of its content (SNF) and fulfillment of all other safety requirements for storage and transportation of SNF. Radiation protection against emissions and non-distribution of activity outside the cask is ensured by physical barriers, i.e. all-metal or composite body, shells, inner cavities for irradiated fuel assemblies (SFA), lids with sealing systems. Residual heat release of SFA is discharged to the environment by natural way: through emission and convection of surrounding air. By now more than 100 dual purpose packaging sets TUK-108/1 are in operation in the mode of interim storage and transportation of SNF from decommissioned nuclear powered submarines (NPS). In accordance with certificate, spent fuel is stored in TUK-108/1 on the premises of plants involved in NPS dismantlement for 2 years, whereupon it is transported for processing to PO Mayak. At one Far Eastern plant Zvezda involved in NPS dismantlement there arose a complicated situation due to necessity to extend period of storage of SNF in TUK- 108/1. To ensure safety over a longer period of storage of SNF in TUK-108/1 it is essential to modify conditions of storage by removing of residual water and filling the inner cavity of the cask with an inert gas. Within implementation of the international 1.1- 2 project Development of drying technology for the cask TUK-108/1 intended for naval SNF under the Program, there has been developed the technology of preparation of the cask for long-term storage of SNF in TUK-108/1, the design of a mobile TUK-108

  17. Design Calculations for Gas Flow & Diffusion Behavior in the Large Diameter Container & Cask

    SciTech Connect

    PIEPHO, M.G.

    2003-11-06

    This report describes the calculations for the gas behavior in the void volumes or gas spaces of the sludge Large Diameter Container (LDC) and Cask. The objective is to prevent flammable gas conditions in the LDC and Cask gas spaces. This is achieved by the Active Inert Ventilation System (AIVS), which uses argon gas for dilution purposes. With AIVS, the oxygen content is kept below 4 to 5 vol% in the LDC, and the hydrogen content is kept below 4 vol% in the Cask before its purge at the KE Basin. After the Cask sweep-through purge with argon at the KE Basin, oxygen stays below 4 to 5% in the LDC until two LDC ports are opened at T Plant. The oxygen content stays below 4% in the Cask until the Cask lid is opened at T Plant. The analysis here assumes that any oxygen generated in the sludge is consumed by the uranium and uranium dioxide (SNF-18133, ''Gas Behavior in Large Diameter Containers (LDCs) During and Following Loading with 105K East Sludge''). Thus, oxygen production from radiolysis is not included in this report, but hydrogen from radiolysis and from chemical reactions between uranium and water are considered, depending on the scenario being analyzed. The analysis starts immediately after the final decant at K Basin, when argon is assumed to be the only gas in the LDC gas space, except for the normal water vapor. The oxygen ingress is calculated during the disconnecting of the lined hoses from the LDC, during the time that air is surrounding the LDC with two NucFil-type filters in place after the disconnect, before the Cask is sealed, and, finally, during the sweep-through Cask purge at the KE Basin. Dissolution of oxygen from water due to increasing sludge temperatures (mainly during hot transport to the T Plant) is also included. The analysis includes the gas behavior during the T-Plant operations, which include the venting after the LDC/Cask are received at T Plant, the Cask sweep-through purge, the LDC purge with forced argon delivery into the LDC with 1

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

    SciTech Connect

    Dunn, K.; Bellamy, S.; Daugherty, W.; Sindelar, R.; Skidmore, E.

    2013-08-18

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

  19. Emissions of greenhouse gases from the use of transportation fuels and electricity

    SciTech Connect

    DeLuchi, M.A. )

    1991-11-01

    This report presents estimates of full fuel-cycle emissions of greenhouse gases from using transportation fuels and electricity. The data cover emissions of carbon dioxide (CO{sub 2}), methane, carbon monoxide, nitrous oxide, nitrogen oxides, and nonmethane organic compounds resulting from the end use of fuels, compression or liquefaction of gaseous transportation fuels, fuel distribution, fuel production, feedstock transport, feedstock recovery, manufacture of motor vehicles, maintenance of transportation systems, manufacture of materials used in major energy facilities, and changes in land use that result from using biomass-derived fuels. The results for electricity use are in grams of CO{sub 2}-equivalent emissions per kilowatt-hour of electricity delivered to end users and cover generating plants powered by coal, oil, natural gas, methanol, biomass, and nuclear energy. The transportation analysis compares CO{sub 2}-equivalent emissions, in grams per mile, from base-case gasoline and diesel fuel cycles with emissions from these alternative- fuel cycles: methanol from coal, natural gas, or wood; compressed or liquefied natural gas; synthetic natural gas from wood; ethanol from corn or wood; liquefied petroleum gas from oil or natural gas; hydrogen from nuclear or solar power; electricity from coal, uranium, oil, natural gas, biomass, or solar energy, used in battery-powered electric vehicles; and hydrogen and methanol used in fuel-cell vehicles.

  20. Emissions of greenhouse gases from the use of transportation fuels and electricity. Volume 1, Main text

    SciTech Connect

    DeLuchi, M.A.

    1991-11-01

    This report presents estimates of full fuel-cycle emissions of greenhouse gases from using transportation fuels and electricity. The data cover emissions of carbon dioxide (CO{sub 2}), methane, carbon monoxide, nitrous oxide, nitrogen oxides, and nonmethane organic compounds resulting from the end use of fuels, compression or liquefaction of gaseous transportation fuels, fuel distribution, fuel production, feedstock transport, feedstock recovery, manufacture of motor vehicles, maintenance of transportation systems, manufacture of materials used in major energy facilities, and changes in land use that result from using biomass-derived fuels. The results for electricity use are in grams of CO{sub 2}-equivalent emissions per kilowatt-hour of electricity delivered to end users and cover generating plants powered by coal, oil, natural gas, methanol, biomass, and nuclear energy. The transportation analysis compares CO{sub 2}-equivalent emissions, in grams per mile, from base-case gasoline and diesel fuel cycles with emissions from these alternative- fuel cycles: methanol from coal, natural gas, or wood; compressed or liquefied natural gas; synthetic natural gas from wood; ethanol from corn or wood; liquefied petroleum gas from oil or natural gas; hydrogen from nuclear or solar power; electricity from coal, uranium, oil, natural gas, biomass, or solar energy, used in battery-powered electric vehicles; and hydrogen and methanol used in fuel-cell vehicles.

  1. Multi-fuel reformers for fuel cells used in transportation: Assessment of hydrogen storage technologies. Phase 1, Final report

    SciTech Connect

    Not Available

    1994-03-01

    This report documents a portion of the work performed Multi-fuel Reformers for Fuel Cells Used in Transportation. One objective for development is to develop advanced fuel processing systems to reform methanol, ethanol, natural gas, and other hydrocarbons into hydrogen for use in transportation fuel cell systems, while a second objective is to develop better systems for on-board hydrogen storage. This report examines techniques and technology available for storage of pure hydrogen on board a vehicle as pure hydrogen of hydrides. The report focuses separately on near- and far-term technologies, with particular emphasis on the former. Development of lighter, more compact near-term storage systems is recommended to enhance competitiveness and simplify fuel cell design. The far-term storage technologies require substantial applied research in order to become serious contenders.

  2. Design Calculations for Gas Flow & Diffusion Behavior in the large Diameter Container & Cask

    SciTech Connect

    PIEPHO, M.G.

    2003-10-21

    This report describes the calculations for the gas behavior in the void volumes or gas spaces of the sludge Large Diameter Container (LDC) and Cask. The objective is to prevent flammable gas conditions in the LDC and Cask gas spaces. This is achieved by the Active Inert Ventilation System (AIVS), which uses argon gas for dilution purposes. With AIVS, the oxygen content is kept below 5 vol% in the LDC, and the hydrogen content is kept below 4 vol% in the Cask before its purge at the KE Basin. After the Cask sweep-through purge with argon at the KE Basin, oxygen is kept below 5% in both the Cask and the LDC. The analysis here assumes that any oxygen generated in the sludge is consumed by the uranium and uranium dioxide (SNF-18133, ''Gas Behavior in Large Diameter Containers (LDCs) During and Following Loading with 105K East Sludge''). Thus, oxygen production from radiolysis is intentionally not included in this report, but hydrogen from radiolysis and from chemical reactions between uranium and water are considered, depending on the scenario being analyzed. The analysis starts immediately after the final decant at K Basin, when argon is assumed to be the only gas in the LDC gas space, except for the normal water vapor. The oxygen ingress is calculated during the disconnecting of the lines hoses from the LDC, during the time that air is surrounding the LDC with two NucFil-type filters in place after the disconnect, before the Cask is sealed, and, finally, during the sweep-through Cask purge. Dissolution of oxygen from water due to increasing sludge temperatures (mainly during hot transport to the T Plant) is also included.

  3. ADVANCED HYDROGEN TRANSPORT MEMBRANES FOR VISION 21 FOSSIL FUEL PLANTS

    SciTech Connect

    Shane E. Roark; Tony F. Sammells; Adam Calihman; Andy Girard; Pamela M. Van Calcar; Richard Mackay; Tom Barton; Sara Rolfe

    2001-01-30

    Eltron Research Inc., and team members CoorsTek, McDermott Technology, Inc., Sued Chemie, Argonne National Laboratory, and Oak Ridge National Laboratory are developing an environmentally benign, inexpensive, and efficient method for separating hydrogen from gas mixtures produced during industrial processes, such as coal gasification. This objective is being pursued using dense membranes based in part on Eltron-patented ceramic materials with a demonstrated ability for proton and electron conduction. The technical goals are being addressed by modifying single-phase and composite membrane composition and microstructure to maximize proton and electron conductivity without loss of material stability. Ultimately, these materials must enable hydrogen separation at practical rates under ambient and high-pressure conditions, without deactivation in the presence of feedstream components such as carbon dioxide, water, and sulfur. This project was motivated by the National Energy Technology Laboratory (NETL) Vision 21 initiative which seeks to economically eliminate environmental concerns associated with the use of fossil fuels. The proposed technology addresses the DOE Vision 21 initiative in two ways. First, this process offers a relatively inexpensive solution for pure hydrogen separation that can be easily incorporated into Vision 21 fossil fuel plants. Second, this process could reduce the cost of hydrogen, which is a clean burning fuel under increasing demand as supporting technologies are developed for hydrogen utilization and storage. Additional motivation for this project arises from the potential of this technology for other applications. Membranes testing during this reporting period were greater than 1 mm thick and had the general perovskite composition AB{sub 1-x}B'{sub x}O{sub 3-{delta}}, where 0.05 {<=} x {<=} 0.3. These materials demonstrated hydrogen separation rates between 1 and 2 mL/min/cm{sup 2}, which represents roughly 20% of the target goal for

  4. Present experience of NRI REZ with preparation of spent nuclear fuel shipment to Russian Federation

    SciTech Connect

    Svitak, F.; Broz, V.; Hrehor, M.; Marek, M.; Novosad, P.; Podlaha, J.; Rychecky, J.

    2008-07-15

    The Nuclear Research Institute Rez plc (NRI) jointed the Russian Research Reactor Fuel Return (RRRFR) programme under the US-Russian Global Threat Reduction Initiative (GTRI) initiative and started the preparation of the spent nuclear fuel (SNF) shipment from the LVR-15 research reactor back to the Russian Federation (RF). The transport of 16 SKODA VPVR/M casks with EK-10, IRT-2M 80 %, and IRT-2M 36% fuel types is planned for the autumn of 2007. The paper describes the experience gained so far during the preparatory works for the SNF shipment (facility equipment modification, cask licenses) and the actual preparation of the SNF for transport, in particular its checking, repacking in a hot cell, loading into the VPVR/M casks, drying, manipulation, completion of the transport documentation, etc., including its transport to the SNF storage facility at the NRI before it is shipped to the RF. The paper also briefly describes a regulatory framework for these activities with a focus on legislative and methodological aspects of the return of vitrified waste back to the Czech Republic. (author)

  5. 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.

  6. A Fuzzy Modeling Approach to Road Transport with Application to a Case of Spent Nuclear Fuel Transport

    SciTech Connect

    Marseguerra, Marzio; Zio, Enrico; Bianchi, Mauro

    2004-06-15

    In this paper, we propose a general fuzzy inference approach to building a model of hazardous road transport that relates given traffic, weather, and vehicle-speed conditions to the accident rate. The development of the model is discussed in detail, and its validation is provided with reference to literature data regarding the transport of spent nuclear fuel to its final confinement repository.

  7. Site Specific Analyses of a Spent Nuclear Fuel Transportation Accident

    SciTech Connect

    Biwer, B. M.; Chen, S. Y.

    2003-02-24

    The number of spent nuclear fuel (SNF) shipments is expected to increase significantly during the time period that the United States' inventory of SNF is sent to a final disposal site. Prior work estimated that the highest accident risks of a SNF shipping campaign to the proposed geologic repository at Yucca Mountain were in the corridor states, such as Illinois. The largest potential human health impacts would be expected to occur in areas with high population densities such as urban settings. Thus, our current study examined the human health impacts from the most plausible severe SNF transportation accidents in the Chicago metropolitan area. The RISKIND 2.0 program was used to model site-specific data for an area where the largest impacts might occur. The results have shown that the radiological human health consequences of a severe SNF rail transportation accident on average might be similar to one year of exposure to natural background radiation for those persons living a nd working in the most affected areas downwind of the actual accident location. For maximally exposed individuals, an exposure similar to about two years of exposure to natural background radiation was estimated. In addition to the accident probabilities being very low (approximately 1 chance in 10,000 or less during the entire shipping campaign), the actual human health impacts are expected to be lower if any of the accidents considered did occur, because the results are dependent on the specific location and weather conditions, such as wind speed and direction, that were selected to maximize the results. Also, comparison of the results of longer duration accident scenarios against U.S. Environmental Protection Agency guidelines was made to demonstrate the usefulness of this site-specific analysis for emergency planning purposes.

  8. The dieselization of America: An integrated strategy for future transportation fuels

    SciTech Connect

    Eberhardt, J.J.

    1997-12-31

    The Diesel Cycle engine has already established itself as the engine-of-choice for the heavy duty transport industry because of its fuel efficiency, durability, and reliability. In addition, it has also been shown to be capable of using alternative fuels, albeit at efficiencies lower than that achieved with petroleum-derived diesel fuel. Alternative fuel dedicated engines have not made significant penetration of the heavy duty truck market because truck fleet operators need a cost-competitive fuel and reliable supply and fueling infrastructure. In lieu of forcing diverse fuels from many diverse domestic feedstocks onto the end-users, the Office of Heavy Vehicle Technologies envisions that a future fuels strategy for the heavy duty transport sector is one where the diverse feedstocks are utilized to provide a single fuel specification (dispensed from the existing fueling infrastructure) that would run efficiently in a single high efficiency energy conversion device, the Diesel Cycle engine. In so doing, the US Commercial transport industry may gain a measure of security from the rapid fuel price increases by relying less on a single feedstock source to meet its increasing fuel requirements.

  9. Radioactive materials shipping cask anticontamination enclosure

    DOEpatents

    Belmonte, Mark S.; Davis, James H.; Williams, David A.

    1982-01-01

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

  10. 77 FR 20438 - Independent Spent Fuel Storage Installation, Virginia Electric and Power Company: North Anna...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-04

    .... (TN) NUHOMS HD Storage System (HD-32PTH) dry storage casks, under the Certificate of Compliance (CoC...: The CoC is the NRC approved design for each dry storage cask system. The proposed action would grant... specifically require storage of spent nuclear fuel under a general license in dry storage casks approved...

  11. Fuel-mix, fuel efficiency, and transport demand affect prospects for biofuels in northern Europe.

    PubMed

    Bright, Ryan M; Strømman, Anders Hammer

    2010-04-01

    Rising greenhouse gas (GHG) emissions in the road transport sector represents a difficult mitigation challenge due to a multitude of intricate factors, namely the dependency on liquid energy carriers and infrastructure lock-in. For this reason, low-carbon renewable energy carriers, particularly second generation biofuels, are often seen as a prominent candidate for realizing reduced emissions and lowered oil dependency over the medium- and long-term horizons. However, the overarching question is whether advanced biofuels can be an environmentally effective mitigation strategy in the face of increasing consumption and resource constraints. Here we develop both biofuel production and road transport consumption scenarios for northern Europe-a region with a vast surplus of forest bioenergy resources-to assess the potential role that forest-based biofuels may play over the medium- and long-term time horizons using an environmentally extended, multiregion input-output model. Through scenarios, we explore how evolving vehicle technologies and consumption patterns will affect the mitigation opportunities afforded by any future supply of forest biofuels. We find that in a scenario involving ambitious biofuel targets, the size of the GHG mitigation wedge attributed to the market supply of biofuels is severely reduced under business-as-usual growth in consumption in the road transport sector. Our results indicate that climate policies targeting the road transport sector which give high emphases to reducing demand (volume), accelerating the deployment of more fuel-efficient vehicles, and promoting altered consumption patterns (structure) can be significantly more effective than those with single emphasis on expanded biofuel supply. PMID:20163088

  12. REDUCING ULTRA-CLEAN TRANSPORTATION FUEL COSTS WITH HYMELT HYDROGEN

    SciTech Connect

    Donald P. Malone; William R. Renner

    2003-07-31

    This report describes activities for the third quarter of work performed under this agreement. Atmospheric testing was conducted as scheduled on June 5 through June 13, 2003. The test results were encouraging, however, the rate of carbon dissolution was below expectations. Additional atmospheric testing is scheduled for the first week of September 2003. Phase I of the work to be done under this agreement consists of conducting atmospheric gasification of coal using the HyMelt technology to produce separate hydrogen rich and carbon monoxide rich product stream. In addition smaller quantities of petroleum coke and a low value refinery stream will be gasified. DOE and EnviRes will evaluate the results of this work to determine the feasibility and desirability of proceeding to Phase II of the work to be done under this agreement, which is gasification of the above-mentioned feeds at a gasifier pressure of approximately 5 bar. The results of this work will be used to evaluate the technical and economic aspects of producing ultra-clean transportation fuels using the HyMelt technology in existing and proposed refinery configurations.

  13. DESIGN EVALUATION OF A LARGE CONCRETE CASK TO MEET IP-2 REQUIREMENTS

    SciTech Connect

    Shappert, L.B.

    2001-08-30

    Oak Ridge National Laboratory (ORNL) has a large quantity of low-level waste, in the form of concrete monoliths, that are stored in large concrete vaults in ORNL's Melton Valley Storage Tanks (MVST). During FY 2000, a number of the monoliths were transferred from the concrete vaults to a Nuclear Regulatory Commission (NRC)-certified lead-shielded cask and shipped to the Nevada Test Site (NTS) for disposal. This activity has resulted in (1) increased radiation exposure both when the monoliths were transferred to the lead-shielded cask and when they were unloaded and buried at the NTS and (2) high cask rental and shipping costs for the program, and (3) the accumulation of empty vaults at ORNL which will also have to be disposed of at NTS, adding a significant additional transportation cost. As a result, Department of Energy (DOE)--Oak Ridge has been exploring ways to ship the MVST cask with its monolith to the NTS for disposal as a unit. To do this, the MVST cask would have to be self-certified as meeting IP-2 package requirements.

  14. Cost Analysis of Fuel Cell Systems for Transportation Compressed Hydrogen and PEM Fuel Cell System

    SciTech Connect

    Eric J. Carlson

    2004-10-20

    PEMFC technology for transportation must be competitive with internal combustion engine powertrains in a number of key metrics, including performance, life, reliability, and cost. Demonstration of PEMFC cost competitiveness has its own challenges because the technology has not been applied to high volume automotive markets. The key stack materials including membranes, electrodes, bipolar plates, and gas diffusion layers have not been produced in automotive volumes to the exacting quality requirements that will be needed for high stack yields and to the evolving property specifications of high performance automotive stacks. Additionally, balance-of-plant components for air, water, and thermal management are being developed to meet the unique requirements of fuel cell systems. To address the question of whether fuel cells will be cost competitive in automotive markets, the DOE has funded this project to assess the high volume production cost of PEM fuel cell systems. In this report a historical perspective of our efforts in assessment of PEMFC cost for DOE is provided along with a more in-depth assessment of the cost of compressed hydrogen storage is provided. Additionally, the hydrogen storage costs were incorporated into a system cost update for 2004. Assessment of cost involves understanding not only material and production costs, but also critical performance metrics, i.e., stack power density and associated catalyst loadings that scale the system components. We will discuss the factors influencing the selection of the system specification (i.e., efficiency, reformate versus direct hydrogen, and power output) and how these have evolved over time. The reported costs reflect internal estimates and feedback from component developers and the car companies. Uncertainty in the cost projection was addressed through sensitivity analyses.

  15. Cask Processing Enclosure Specification/Operation - 12231

    SciTech Connect

    Gentry, Ronald

    2012-07-01

    Following an evaluation of throughput rates in the Hot Cell at the Transuranic Waste Processing Center and considering the variability in the waste with respect to actual dose rates a new approach to processing transuranic waste was necessary. Compounding the issue was the remote equipment poor reliability and high down-time. After considering all the factors, the evaluation resulted in the design and construction of a new waste processing area for handling the concrete casks that predominately contain contact handled transuranic (TRU) waste. The area is called the Cask Processing Enclosure and essentially the Cask Processing Enclosure mimics the projects current process techniques used for processing Contact Handled -TRU waste in the existing Box Breakdown Area and Glovebox. The Cask Processing Enclosure approach was developed based on a review of the RH processing throughput rates in the Hot Cell. As the process was reviewed consideration was given to the variability in the waste with respect to actual dose rates and the lack of equipment reliability and high wear in the Hot Cell. Based on that review, a new contact handled processing area for handling the concrete casks is being constructed and startup is expected shortly following WM2012. The Cask Processing Enclosure essentially mimics the projects current process techniques used for processing Contact Handled waste in the existing Box Breakdown Area and Glovebox and the design takes into consideration six years of operational experience. (authors)

  16. Future long-range transports - Prospects for improved fuel efficiency

    NASA Technical Reports Server (NTRS)

    Nagel, A. L.; Alford, W. J., Jr.; Dugan, J. F., Jr.

    1975-01-01

    A status report is provided on current thinking concerning potential improvements in fuel efficiency and possible alternate fuels. Topics reviewed are: historical trends in airplane efficiency; technological opportunities including supercritical aerodynamics, vortex diffusers, composite materials, propulsion systems, active controls, and terminal-area operations; unconventional design concepts, and hydrogen-fueled airplane.

  17. Future long-range transports: Prospects for improved fuel efficiency

    NASA Technical Reports Server (NTRS)

    Nagel, A. L.; Alford, W. J., Jr.; Dugan, J. F., Jr.

    1975-01-01

    A status report is provided on current thinking concerning potential improvements in fuel efficiency and possible alternate fuels. Topics reviewed are: (1) historical trends in airplane efficiency; (2) technological opportunities including supercritical aerodynamics, (3) vortex diffusers, (4) composite materials, (5) propulsion systems, (6) active controls, and terminal-area operations; (7) unconventional design concepts, and (8) hydrogen-fueled airplane.

  18. Fire tests and analyses of a rail cask-sized calorimeter.

    SciTech Connect

    Figueroa, Victor G.; Lopez, Carlos; Suo-Anttila, Ahti Jorma; Greiner, Miles

    2010-10-01

    Three large open pool fire experiments involving a calorimeter the size of a spent fuel rail cask were conducted at Sandia National Laboratories Lurance Canyon Burn Site. These experiments were performed to study the heat transfer between a very large fire and a large cask-like object. In all of the tests, the calorimeter was located at the center of a 7.93-meter diameter fuel pan, elevated 1 meter above the fuel pool. The relative pool size and positioning of the calorimeter conformed to the required positioning of a package undergoing certification fire testing. Approximately 2000 gallons of JP-8 aviation fuel were used in each test. The first two tests had relatively light winds and lasted 40 minutes, while the third had stronger winds and consumed the fuel in 25 minutes. Wind speed and direction, calorimeter temperature, fire envelop temperature, vertical gas plume speed, and radiant heat flux near the calorimeter were measured at several locations in all tests. Fuel regression rate data was also acquired. The experimental setup and certain fire characteristics that were observed during the test are described in this paper. Results from three-dimensional fire simulations performed with the Cask Analysis Fire Environment (CAFE) fire code are also presented. Comparisons of the thermal response of the calorimeter as measured in each test to the results obtained from the CAFE simulations are presented and discussed.

  19. Driving it home: choosing the right path for fueling North America's transportation future

    SciTech Connect

    Ann Bordetsky; Susan Casey-Lefkowitz; Deron Lovaas; Elizabeth Martin-Perera; Melanie Nakagawa; Bob Randall; Dan Woynillowicz

    2007-06-15

    North America faces an energy crossroads. With the world fast approaching the end of cheap, plentiful conventional oil, we must choose between developing ever-dirtier sources of fossil fuels -- at great cost to our health and environment -- or setting a course for a more sustainable energy future of clean, renewable fuels. This report explores the full scale of the damage done by attempts to extract oil from liquid coal, oil shale, and tar sands; examines the risks for investors of gambling on these dirty fuel sources; and lays out solutions for guiding us toward a cleaner fuel future. Table of contents: Executive Summary; Chapter 1: Transportation Fuel at a Crossroads; Chapter 2: Canadian Tar Sands: Scraping the Bottom of the Barrel in Endangered Forests; Chapter 3: Oil Shale Extraction: Drilling Through the American West; Chapter 4: Liquid Coal: A 'Clean Fuel' Mirage; Chapter 5: The Investment Landscape: Dirty Fuels Are Risky Business; Chapter 6: The Clean Path for Transportation and Conclusion.

  20. Preliminary assessment of costs and risks of transporting spent fuel by barge

    SciTech Connect

    Tobin, R.L.; Meshkov, N.K.; Jones, R.H.

    1985-12-01

    The purpose of this study is to analyze the costs and risks associated with transporting spent fuel by barge. The barge movements would be made in combination with rail movements to transport spent fuel from plants to a repository. For the purpose of this analysis, three candidate repository sites are analyzed: Yucca Mountain, Nevada, Deaf Smith, Texas, and Hanford, Washington. This report complements a report prepared by Sandia National Laboratories in 1984 that analyzes the costs and risks of transporting spent fuel by rail and by truck to nine candidate repository sites.