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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. Experience with the loading and transport of fuel assembly transport casks, including CASTOR casks, and the radiation exposure of personnel.

    PubMed

    Bentele, W; Kinzelmann, T

    1999-12-01

    In 1997 and 1998, six spent fuel assembly transports started from the nuclear power plant Gemeinschaftskernkraftwerk Neckar (GKN), using CASTOR-V19 casks. Professor Kuni of Marburg University challenged the statement made by the German Federal Office for Radiation Protection (Bundesamt für Strahlenschutz (BfS)) based on accepted scientific knowledge, according to which so-called CASTOR transports present no risk, either to the population or to the escorting police units. This paper shows that the collective dose during the loading of the CASTOR casks amounted to 4.5 mSv (gamma and neutrons) per cask at the most, and that the maximum individual dose amounted to 0.26 mSv. In addition to these doses, the collective dose during handling and transport must be considered: this amounted to 0.35 mSv (gamma and neutrons). The dose to the police escort was <0.1 mSv (gamma and neutrons). In the light of these circumstances, this report is presented on contamination determined during the transport of CASTOR casks and of other spent fuel casks. The controls of spent fuel transports carried out since 1978, mainly with NTL 11 spent fuel casks, revealed that about one fifth of the transport casks which left the GKN with a surface contamination of <4 Bq cm(-2) (limit for surface contamination), presented degrees of contamination >4 Bq cm(-2) upon reaching the Valognes/Cogema terminal. However, transport casks coming from French plants also revealed degrees of contamination >4 Bq cm(-2), as well as 'hot spots'. No such contamination was found on NTL 11 casks transported from the GKN to Sellafield. Neither was any increased contamination found upon the arrival of CASTOR-V19 casks transported from GKN to Gorleben or Ahaus. The partially sensationalist media reports were inversely proportional to the actual radiological relevance of the matter. The German Commission on Radiation Protection (SSK) confirmed that the radiological effect of such contaminated spent fuel transports is

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

    SciTech Connect

    Shah, M. J.; Klymyshyn, Nicholas A.; Adkins, Harold E.; Koeppel, Brian J.

    2007-03-30

    The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 [1]. Both the International Atomic Energy Agency regulations for transporting radioactive materials [2, paragraph 727], and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 [3], for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask [3] for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS® [4] and LS DYNA™ [5] finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event.

  11. 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.

  12. 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.

  13. 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.

  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. HI-STAR 100 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

    SciTech Connect

    Shah, M. J.; Klymyshyn, Nicholas A.; Adkins, Harold E.; Koeppel, Brian J.

    2007-03-30

    The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (CFR), Title 10, Part 71 [1]. Both the International Atomic Energy Agency (IAEA) regulations for transporting radioactive materials [2, paragraph 727], and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop impact event on a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC-certified transportation casks, the HI-STAR 100 [3], for drop impact events. The specific area examined is the behavior of the bolted connections in the “overpack” top flange and the closure plate, which are significantly loaded during the hypothetical drop impact event. The term “overpack” refers to the cask that receives and contains a sealed multi-purpose canister (MPC) containing spent nuclear fuel. The analytical work to evaluate the NRC-certified HI-STAR 100 spent fuel transport cask [3] for a 9-m (30-ft) drop impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS® [4] and LS DYNA™ [5] finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free-drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the closure plate surfaces may occur, the seals would

  16. 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.

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

    SciTech Connect

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

    2014-01-01

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

  18. Neutron spectra and dose-rate measurements around a transport cask for spent reactor fuel

    NASA Astrophysics Data System (ADS)

    Rimpler, Arndt

    1997-02-01

    A storage facility with a capacity of 420 containers is available for the interim storage of spent fuel from power reactors at the village of Gorleben in Germany. During transportation and storage of spent fuel casks radiation exposure of the personnel is dominated by neutrons. The routine control of the dose rate limits according to the transport regulations and the licence of the storage facility is performed with conventional neutron survey meters. These monitors, calibrated for fast neutrons at radionuclide neutron sources, usually overestimate the real dose rate in unknown neutron fields. In this paper, a series of measurements with several monitoring instruments near a transport cask of the CASTOR type is presented. The results are compared with reference data for the does equivalents calculated from the measured fluence spectra using a Bonner multisphere spectrometer. Besides reliable information about neutron spectra and dose rates at the container, it was found that some of the rem counters overestimate the true dose rate by a factor of 2 or more.

  19. 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.

  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. 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.

  2. A new shielding analysis method for used fuel transport and storage casks

    NASA Astrophysics Data System (ADS)

    Léger, Vincent; Kitsos, Stavros

    2017-09-01

    To provide a cask with the largest possible loading capacity of spent fuel assemblies with the largest practicable burnup and shortest cooling time within all safety requirements, AREVA TN has adapted its design process and developed a more elaborated shielding analysis method. Taking advantage of the potential heterogeneities between sources of fuel assemblies to be loaded, and the self-shielding of assemblies loaded at the basket centre by the assemblies loaded at the basket periphery, the result of this method is expressed under the shape of a linear inequalities system allowing to optimize the cask capacity and performance.

  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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  4. 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.

  5. 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)

  6. Feedback on the use of the MX6 Mox Fuel transport cask: reduction of the dose uptake during operations

    SciTech Connect

    Blachet, L.; Lallemant, Th.

    2007-07-01

    In the framework of the quality, safety and environment policy of AREVA, TN International has implemented a global management system according to ISO 9001, OHSAS 18001 and ISO 14001 requirements with certification obtained from third part organization (1). The design of the MX6 cask is an example of the implementation of this system in order to guarantee safety and the health of everyone involved and the protection of the environment. The MX6 design has allowed ALARA dose rates for the workers during all the phases of use of the cask, to be significantly reduced compared to previous design. The MX6 cask was developed by TN International for the transport of either BWR or PWR fresh MOX fuel assemblies. Replacing the previous SIEMENS type III and SIEMENS BWR packaging, the MX6 has been firstly used in the German Nuclear Power Plants. Complying with the TS-R-1 (IAEA 1996) regulations, the MX6 cask is based on innovative solutions implemented at each step of the design and the manufacturing. Its design includes an efficient neutron shielding for high Plutonium content and an easy use content restraining system. The large payload of the MX6 cask, 6 PWR MOX fuel assemblies or 16 BWR MOX fuel assemblies, minimizes the doses uptake during its unloading at the NPP. Moreover, new sequences of loading and unloading operations were proposed for testing and implementation in each Nuclear Facility. Concurrently, total dose uptakes by the operators were assessed in order to prove the efficiency of the packaging and the proposed sequences. In this paper, the main contributors to the transports to Germany with the MX6 cask will present their involvement and feedback for the reduction of the dose uptakes by the operators during the loading and unloading operations. Presently in use at GUNDREMMINGEN and ISAR Nuclear Power Plants (NPPs), the MX6 cask use will be extended to other German and Swiss NPPs from 2006 onwards. (1) AFAQ-AFNOR Certification for ISO 9001, OHSAS 18001 and ISO

  7. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    SciTech Connect

    Duckett, A.J.; Sorenson, K.B.

    1985-09-01

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref.

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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.

  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. Transportation cask contamination weeping: A program leading to prevention

    SciTech Connect

    Bennett, P. C.; Doughty, D. H.; Chambers, W. B.

    1991-01-01

    This paper describes the problem of cask contamination weeping, and efforts to understand the phenomenon and to eliminate its occurrence during spent nuclear fuel transport. The paper analyses of field experience and scoping experiments, and concentrates on current modelling and experimental validation efforts. The weeping'' phenomenon associated with spent fuel transportation casks (also known as sweating'') is believed to be due to the conversion of fixed contamination on the external surface of the cask to a removable form. Spent fuel transportation casks are loaded under water at nuclear power plants in a spent fuel storage pool, exposing the cask surfaces to contamination by radionuclides present in the pool water including {sup 137}Cs, {sup 134}Cs, and {sup 60}Co. The external surfaces of loaded casks are routinely surveyed for removable contamination and decontaminated to 1/10 of the US and IAEA regulatory limits prior to being released for shipment (49CFR 1983, IAEA 1989). However, 3% to 8% of US spent fuel casks have arrived at final destinations with removable surface contamination in excess of that allowed by regulation, though many preshipment surveys have shown contaminant levels to be within allowable limits (Grella 1987). Attempts to reduce the incidence of weeping have met with limited success and resulted in time-consuming operational constraints and procedures that significantly increase cask processing times and occupational exposures at loading facilities. As the US Department of Energy (DOE) moves toward a high volume spent fuel transportation campaign beginning in 1998, the elimination of weeping occurrence and minimization of operational constraints has received increased attention.

  15. 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.

  16. 78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-26

    ... RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9... amends the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 Cask...

  17. 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.

  18. 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.

  19. 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.

  20. Cask performance and interface specifications for shipment of US spent light water reactor fuel

    SciTech Connect

    Sanders, T.L.; Allen, G.C.; Wilmot, E.L.

    1986-01-01

    Preliminary cask physical performance specifications and interface guidelines have been defined to support the development of a family of casks and transporters for shipments in the United States (US) of spent light water reactor fuel. These shipments will be made from US commercial reactor facilities to high-level waste receiving facilities. The specified hardware consists of both truck and rail/barge casks along with their associated transporters.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-26

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI... revising the Holtec International HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel...

  12. 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

  13. 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.

  14. 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

    ... 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY... (NRC) is amending its spent fuel storage regulations by revising the Holtec International HI- STORM 100... the HI-STORM 100U part of the HI-STORM 100 Cask System and updates the thermal model and methodology...

  15. COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

    DOE PAGES

    Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

    2017-09-01

    COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

  16. 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

  17. 78 FR 73456 - 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

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI... regulations by revising the Holtec International HI-STORM 100 Cask System listing within the ``List of.... Amendment No. 9 broadens the subgrade requirements for the HI-STORM 100U part of the HI-STORM 100 Cask...

  18. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    SciTech Connect

    Rauch, Eric Benton

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each cask's neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  19. 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.

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

    SciTech Connect

    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. 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.

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

    SciTech Connect

    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. 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.

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

    DOE PAGES

    Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; ...

    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

  3. 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.

  4. Use of transportable storage casks in the nuclear waste management system

    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. A summary of the results of cost estimates developed in Section 4.0 and the Appendices to this report is shown in Tables 2-1 and 2-2 for instances in which the TSC or SOC were delivered to DOE containing intact fuel assemblies and cans of consolidated fuel, respectively. 2 figs., 14 tabs.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Benchmarking Data for the Proposed Signature of Used Fuel Casks

    SciTech Connect

    Rauch, Eric Benton

    2016-09-23

    A set of benchmarking measurements to test facets of the proposed extended storage signature was conducted on May 17, 2016. The measurements were designed to test the overall concept of how the proposed signature can be used to identify a used fuel cask based only on the distribution of neutron sources within the cask. To simulate the distribution, 4 Cf-252 sources were chosen and arranged on a 3x3 grid in 3 different patterns and raw neutron totals counts were taken at 6 locations around the grid. This is a very simplified test of the typical geometry studied previously in simulation with simulated used nuclear fuel.

  14. 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.

  15. 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.

  16. 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).

  17. 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)

  18. 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.

  19. 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

    ... 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear... the NRC's spent fuel storage regulations to add the Holtec HI-STORM Flood/Wind cask system to the... Holtec HI- STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks'' as...

  20. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    SciTech Connect

    C.E. Sanders

    2005-04-26

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

  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. 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.

  3. 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.

  4. 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 action...

  5. 78 FR 22411 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No.... (Holtec) HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to... technical specifications (TS) and the NRC's Safety Evaluation Report (SER) for the Holtec HI-STORM 100 Cask...

  6. 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

    ... 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear... the Commission) is amending its regulations to add the HI-STORM Flood/Wind cask system to the ``List... spent fuel storage cask designs. Discussion This rule will add the Holtec HI-STORM Flood/Wind (FW) cask...

  7. 78 FR 32077 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-29

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks: MAGNASTOR System AGENCY... Commission (NRC) is withdrawing a direct final rule that would have revised its spent fuel storage... ``List of Approved Spent Fuel Storage Casks.'' The NRC is taking this action because it has received...

  8. 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

    ... 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear... amends the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 System... International HI-STORM 100 System listing within the ``List of Approved Spent Fuel Storage Casks'' to include...

  9. 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.

  10. 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

  11. Testing and COBRA-SFS analysis of the VSC-17 ventilated concrete, spent fuel storage cask

    SciTech Connect

    McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

    1992-04-01

    A performance test of a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask loaded with 17 canisters of consolidated PWR spent fuel generating approximately 15 kW was conducted. The performance test included measuring the cask surface, concrete, air channel surface, and fuel temperatures, as well as cask surface gamma and neutron dose rates. Testing was performed using vacuum, nitrogen, and helium backfill environments. Pretest predictions of cask thermal performance were made using the COBRA-SFS computer code. Analysis results were within 15{degrees}C of measured peak fuel temperature. Peak fuel temperature for normal operation was 321{degrees}C. In general, the surface dose rates were less than 30 mrem/h on the side of the cask and 40 mrem/h on the top of the cask.

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

    SciTech Connect

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

    2003-08-17

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

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

    SciTech Connect

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

    2003-03-17

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

  14. Seismic tipping analysis of a spent nuclear fuel shipping cask sitting on a crush pad

    SciTech Connect

    Uldrich, E.D.; Hawkes, B.D.

    1998-04-01

    A crush pad has been designed and analyzed to absorb the kinetic energy of an accidentally dropped spent nuclear fuel shipping cask into a 44 ft. deep cask unloading pool. Conventional analysis techniques available for evaluating a cask for tipping due to lateral seismic forces assume that the cask rests on a rigid surface. In this analysis, the cask (110 tons) sits on a stainless steel encased (0.25 in. top plate), polyurethane foam (4 ft. thick) crush pad. As the cask tends to rock due to horizontal seismic forces, the contact area between the cask and the crush pad is reduced, increasing the bearing stress, and causing the pivoting corner of the cask to depress into the crush pad. As the crush pad depresses under the cask corner, the pivot point shifts from the corner toward the cask center, which facilitates rocking and potential tipping of the cask. Subsequent rocking of the cask may deepen the depression, further contributing to the likelihood of cask tip over. However, as the depression is created, the crush pad is absorbing energy from the rocking cask. Potential tip over of the cask was evaluated by performing a non-linear, dynamic, finite element analysis with acceleration time history input. This time history analysis captured the effect of a deforming crush pad, and also eliminated conservatisms of the conventional approaches. For comparison purposes, this analysis was also performed with the cask sitting on a solid stainless steel crush pad. Results indicate that the conventional methods are quite conservative relative to the more exacting time history analysis. They also indicate that the rocking motion is less on the foam crush pad than on the solid stainless steel pad.

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

    SciTech Connect

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

    2016-04-29

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

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

    SciTech Connect

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

    2016-04-29

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

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

    SciTech Connect

    Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher; Bacon, Jeffrey Darnell; Plaud-Ramos, Kenie Omar; Morley, Deborah Jean; Hecht, Adam A.

    2016-10-22

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

  18. 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 HI...

  19. 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.

  20. Sensitivity Analysis Applied to the Validation of the 10 B Capture Reaction in Nuclear Fuel Casks

    SciTech Connect

    Goluoglu, S

    2004-03-18

    Boron has commonly been used in nuclear fuel casks to ensure a sufficient margin of subcriticality. The amount of boron used in most casks far exceeds the amount of boron present in any of the available benchmark experiments. Such heavy loadings of boron in the casks may result in considerable spectral differences as compared to the benchmarks, resulting in boron sensitivities that are very different from those of the benchmarks. Before the calculations to determine the nuclear safety margin for various fuel loadings are deemed acceptable, as part of the safety basis, the computer code and cross sections must be validated against experimental benchmarks that cover the area of applicability of the proposed cask design. Therefore, this study was performed to determine if these available benchmarks can be used to validate a criticality code and neutron cross sections for the fuel casks. The sensitivity/uncertainty methodology has been applied to several application cask systems with different boron areal densities. Although, the sensitivities of the nuclear fuel cask applications are not completely covered by the set of benchmarks that were used in this study with regard to the 10B capture cross section, the effect of this lack of coverage on the keff is minimal. Thus, the experimental biases are determined to be appropriate for the cask systems, and no additional bias (penalty) due to high boron loading need be imposed.

  1. 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... International. SAR Title: Final Safety Analysis Report for the HI-STAR 100 Cask System. Docket Number:...

  2. 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... International. SAR Title: Final Safety Analysis Report for the HI-STAR 100 Cask System. Docket Number:...

  3. 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.

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

    PubMed

    Rezaeian, M; Kamali, J

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B4C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

  5. Analysis of a hypothetical dropped spent nuclear fuel shipping cask impacting a floor mounted crush pad

    SciTech Connect

    Hawkes, B.D.; Uldrich, E.D.

    1998-03-01

    A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped spent nuclear fuel shipping cask into a 44-ft. deep cask unloading pool at the Idaho Chemical Processing Plant. The 110-ton Large Cell Cask was assumed to be accidentally dropped onto the parapet of the unloading pool, causing the cask to tumble through the pool water and impact the floor mounted crush pad with the cask`s top corner. The crush pad contains rigid polyurethane foam, which was modeled in a separate computer analysis to simulate the manufacturer`s testing of the foam and to determine the foam`s stress and strain characteristics. This computer analysis verified that the foam was accurately represented in the analysis to follow. A detailed non-linear, dynamic finite element analysis was then performed on the crush pad and adjacent pool structure to assure that a drop of this massive cask does not result in unacceptable damage to the storage facility. Additionally, verification was made that the crush pad adequately protects the cask from severe impact loading. At impact, the cask has significant vertical, horizontal and rotational velocities. The crush pad absorbs much of the energy of the cask through plastic deformation during primary and secondary impacts. After the primary impact with the crush pad, the cask still has sufficient energy to rebound and rotate until it impacts the pool wall. An assessment is made of the damage to the crush pad and pool wall and of the impact loading on the cask.

  6. 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.

  7. 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.

  8. 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.

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

    DOE PAGES

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

    2016-04-29

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

  10. 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.

  11. 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%.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

    ... RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear... Commission) is amending its spent fuel storage regulations by revising the Holtec International HI-STORM 100... and safety will be adequately protected. This direct final rule revises the HI-STORM 100 listing in 10...

  20. 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...

  1. 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...

  2. 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...

  3. 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 C...

  4. 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 C...

  5. 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...

  6. 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...

  7. 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 C...

  8. 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 C...

  9. 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 REACTOR...

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

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 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...

  11. 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...

  12. 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.

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

    DOE PAGES

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

    2017-10-22

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

  14. 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.

  15. Fuel removal, transport, and storage

    SciTech Connect

    Reno, H.W.

    1986-01-01

    The March 1979 accident at Unit 2 of the Three Mile Island Nuclear Power Station (TMI-2) which damaged the core of the reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing the core debris from the reactor, packaging it into canisters, loading canisters into a rail cask, and transporting the debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights how some challenges were resolved, including lessons learned and benefits derived therefrom. Key to some success at TMI was designing, testing, fabricating, and licensing two rail casks, which each provide double containment of the damaged fuel. 10 refs., 12 figs.

  16. Thermal Modeling of a Vertical Dry Storage Cask for Used Nuclear Fuel

    SciTech Connect

    Li, Jie; Liu, Y.

    2016-05-01

    Thermal modeling of temperature profiles of dry casks has been identified as a high-priority item in a U.S. Department of Energy gap analysis. In this work, a three-dimensional model of a vertical dry cask has been constructed for computer simulation by using the ANSYS/FLUENT code. The vertical storage cask contains a welded canister for 32 Pressurized Water Reactor (PWR) used-fuel assemblies with a total decay heat load of 34 kW. To simplify thermal calculations, an effective thermal conductivity model for a 17 x 17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. The effects of canister fill gas (helium or nitrogen), internal pressure (1-6 atm), and basket material (stainless steel or aluminum alloy) were studied to determine the peak cladding temperature (PCT) and the canister surface temperatures (CSTs). The results showed that high thermal conductivity of the basket material greatly enhances heat transfer and reduces the PCT. The results also showed that natural convection affects both PCT and the CST profile, while the latter depends strongly on the type of fill gas and canister internal pressure. Of particular interest to condition and performance monitoring is the identification of canister locations where significant temperature change occurs after a canister is breached and the fill gas changes from high-pressure helium to ambient air. This study provided insight on the thermal performance of a vertical storage cask containing high-burnup fuel, and helped advance the concept of monitoring CSTs as a means to detect helium leakage from a welded canister. The effects of blockage of air inlet vents on the cask's thermal performance were studied. The simulation were validated by comparing the results against data obtained from the temperature measurements of a commercial cask.

  17. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR... storage cask must be designed to provide adequate heat removal capacity without active cooling systems. (g... ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could...

  18. 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.

  19. 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

    ... undamaged and damaged fuel assemblies; minor design modifications to the Vertical Concrete Cask (VCC... as low as is reasonably achievable (ALARA) principles; an increase in the concrete pad compression... in the concrete pad compression strength from 4,000 psi to 6,000 psi; (6) added justification for...

  20. 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)

  1. 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.

  2. 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.

  3. 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.

  4. Assessment of LMFBR spent fuel shipping cask concepts for the CRBRP and the US conceptual design study

    SciTech Connect

    Pope, R B; Ortman, J M; Eakes, R G; Leisher, W B; Dupree, S A

    1980-01-01

    Study of conceptual shipping systems for CRBRP and CDS spent fuel has shown that systems significantly different from those used for LWR spent fuel will be required. In the conceptual design, liquid sodium was assumed to be the coolant in canisters containing the spent fuel assemblies, and multiple levels of containment were provided by canisters, an inner cask lid and an outer cask lid. Cask cooling at the reactor site during loading, and cooldown at the receiving site prior to unloading are significant but tractable problems.

  5. 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

    assemblies inside the welded steel canisters, measurements were performed on the canisters using the Spent Fuel Attribute Monitor (SPAM), which was a neutron coincidence counter very similar in design to the SFCC, but with a larger operational volume to accommodate the canisters. The analysis of the neutron coincidence data from the measurements of the welded steel canisters using SP AM yielded a Canister Attribute (CA) that was shown to be proportional to the total amount of Pu mass that was present inside the canister and comparable to the CA values that were calculated based on the previous SFCC measurements. The next step in securing the spent fuel from the BN-350 reactor is the packaging of the steel canisters into Dual-Use Casks (DUCs) and the transportation of the DUCs to a remote dry storage site.

  6. Thermal evaluation of alternative shipping cask for irradiated experiments

    DOE PAGES

    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

  7. Storage, transportation and disposal system for used nuclear fuel assemblies

    DOEpatents

    Scaglione, John M.; Wagner, John C.

    2017-07-11

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  8. Storage, transportation and disposal system for used nuclear fuel assemblies

    DOEpatents

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  9. 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

    ... U.S. Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising... Environmental Management Programs, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301... U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and...

  10. 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.

  11. 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.

  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. 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.

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

    SciTech Connect

    McCreery, Paul N.

    1980-10-01

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

  15. Packaging and transportation system for K-Basin spent fuel

    SciTech Connect

    Kee, A.T.

    1998-03-03

    This paper describes the cask/transportation system that was designed, procured and delivered to the Hanford K-Basin site at Richland, Washington. The performance requirements and design of the various components -- cask, trailer with cask tie-down system, and the cask operation equipment for the load-out pit -- will be discussed. The presentation will include the details of the factory acceptance testing and its results. The performance requirements for the cask/transportation system was dictated by the constraints imposed by the large number of high priority shipments and the spent fuel pool environment, and the complex interface requirements with other equipment and facility designs. The results of the testing form the basis for the conclusion that the system satisfies the site performance requirements. The cask/transportation system design was driven by the existing facility constraints and the limitations imposed by the large number of shipments over a short two-year period. This system may be useful information for other DOE facilities that may be or will be in a similar situation.

  16. Used Fuel Testing Transportation Model

    SciTech Connect

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

    2014-09-25

    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.

  17. Comparison of spent-fuel cask radiation doses calculated by one- and two-dimensional shielding codes

    SciTech Connect

    Carbajo, J.J. )

    1992-01-01

    Spent-fuel cask shield design and calculation of radiation doses are major parts of the overall cask design. This paper compares radiation doses calculated by one- and two-dimensional or three-dimensional shielding codes. The paper also investigates the appropriateness of using one-dimensional codes for two-dimensional geometries. From these results, it can be concluded that the one-dimensional XSDRNPM/XSDOSE codes are adequate for both radial and axial shielding calculations if appropriate bucklings are used. For radial calculations, no buckling or a buckling equal to the length of the fuel are appropriate. For axial calculations, a buckling at least equal to the diameter of the cask must be used for neutron doses. For gamma axial doses, a buckling around the diameter of the fuel region is adequate. More complicated two- or three-dimensional codes are not needed for these types of problems.

  18. Evaluation of the /sup 252/Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

    SciTech Connect

    Mihalczo, J.T.

    1987-01-01

    The /sup 252/Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt light water reactor (LWR) fuel submerged in fuel storage pools, fully loaded and as they are being loaded. Measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This, in turn, could lead to more cost-effective cask designs. Evaluation of the method for this application was based on experiments already completed at a critical experiments facility using arrays of pressurized water reactor (PWR) fuel pins typical of the size of storage cask configurations, the existence of neutron detectors that can function in shipping cask environments, and the ability to construct ionization chambers containing /sup 252/Cf of adequate intensity for these measurements.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. Summary of the technical review of the safety analysis reports for packaging (SARP) for the transnuclear transport/storage casks: TN-BRP and TN-REG

    SciTech Connect

    Not Available

    1986-07-01

    The Safety Analysis Reports for Packaging for two spent fuel shipping casks were technically reviewed by the Oak Ridge National Laboratory. The casks were designed by Transnuclear, Inc., for shipment of 85 Big Rock Point boiling water reactor fuel elements and 40 R.E. Ginna pressurized water reactor fuel elements from West Valley, New York, to Idaho Falls, Idaho. The intent of the review was to ensure compliance of the casks with the requirements the applicable Federal Regulations contained in 10 CFR Pt. 71 and allow issuance of Department of Energy Certificates of Compliance for transport by the Department of Energy Idaho Operations Office. The review was performed by a team of Oak Ridge National Laboratory staff assembled for their expertise in criticality analysis, shielding, metallurgy, nondestructive testing, thermal analysis, structural analysis, and containment. This report describes the review processes, the findings in each technical area, and the overall conclusion that a Certificate of Compliance could be issued for the proposed single shipment under the specified conditions and constraints.

  8. Transfer Cask Assembly Onsite Transfer of K-Basin Fuel at Hanford

    SciTech Connect

    BRUBAKER, M.L.

    2002-07-01

    Packaging Technology, Inc. was awarded a contract for the design and fabrication of a unique 18.5-ton transportation system to be used on-site at Hanford, WA. Unique aspects include rectangular cask geometry, solid 7 inch-thick stainless steel construction, hinged lid with remotely operable quick release locking mechanism. The lid maintains shielding and material confinement integrity after hypothetical accident conditions, and is equipped with a hydrogen venting system. This paper presents how the project requirements were translated into a successful design, and includes a brief discussion of a confirmation test program.

  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. 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).

  11. 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.

  12. 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/

  13. Compton Dry-Cask Imaging System

    ScienceCinema

    None

    2016-07-12

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

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

  15. Summary report on optimized designs for shipping casks containing 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect

    Bucholz, J.A.

    1983-04-01

    The purpose of this study was to develop new conceptual designs for large Pb, Fe, and U-shielded spent fuel casks which have been optimized for the shipment of 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel assemblies. Design specifications for about 100 cases of potential interest are presented along with a brief 20-page synopsis of the associated analyses. Optimized shielding requirements are presented for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. With respect to criticality, 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. Steady-state and transient heat transfer analyses for casks under nominal and accident conditions were performed using the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. Based on criticality, shielding, and heat transfer considerations, it appears that optimized cask designs could be developed to carry 15 to 18 five-year-old PWR fuel assemblies or as many as 18 to 21 ten-year-old PWR fuel assemblies. 4 figures, 4 tables.

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

    SciTech Connect

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

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

  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. 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.

  19. 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

  20. Shielding analysis of the TRUPACT-series casks for transportation of Hanford HLW

    SciTech Connect

    Banjac, V.; Sanchez, P.E.; Hills, C.R.; Heger, A.S. )

    1993-01-01

    In this paper, the authors propose the possibility of utilizing the TRUPACT-series casks for the transportation of high-level waste (HLW) from the Hanford reservation. The configurations of the TRUPACT series are a rectangular parallelepiped and a right circular cylinder, which are the TRUPACT-1 and -11, respectively. The TRUPACT series was designed as a type B contact-handled transuranic (CH-TRU) waste transportation system for use in Waste Isolation Pilot Plant-related operations and was subjected to type B container accident tests, which it successfully passed. Thus from a safety standpoint, the TRUPACT series is provided with double containment, impact limitation, and fire-retardant capabilities. However, the shielding analysis has shown the major modifications are required to allow for the transport of even a reasonable fraction of Hanford HLW.

  1. ALARA assessment of spent fuel and nuclear waste transportation systems

    SciTech Connect

    Sutherland, S. H.

    1980-01-01

    The effects of ALARA (as low as reasonably achievable) on transportation system costs were evaluated for LWR spent fuel, high-level commercial and defense wastes, and remotely handled TRU waste. Three dose rate specifications were used: 10 mrem/h at 2m, 5 mrem/h, and 2 mrem/h. The evaluation was done for wastes and LWR spent fuel 1, 3, 5, and 10 years old. Gamma shield materials were depleted uranium, lead, and steel; the neutron shield material was water. Results for a 7-element PWR cask show that uranium shielding is the lightest, and that the increased weight of the low dose rate casks results in 1 to 2 million dollars increase in lifetime transportation costs. 6 figures, 3 tables. (DLC)

  2. The thermal analysis of BR-100: A barge/rail nuclear spent fuel transportation container

    SciTech Connect

    Copsey, A.B.

    1992-08-01

    B&W Fuel Company is designing a spent-fuel container called BR-100 that can be used for either barge or rail transport. This paper presents the thermal design and analysis. Both normal operation and hypothetical accident thermal transient conditions are evaluated. The BR-100 cask has a concrete layer than contains free water. During a hypothetical accident, the free water vaporizes and flows from the cask, removing a significant amount of thermal transient energy. The BR-100 transportation package meets the thermal requirements of 10CFR71. It additionally offers substantial margins to established material temperature limits.

  3. The thermal analysis of BR-100: A barge/rail nuclear spent fuel transportation container

    SciTech Connect

    Copsey, A.B.

    1992-01-01

    B W Fuel Company is designing a spent-fuel container called BR-100 that can be used for either barge or rail transport. This paper presents the thermal design and analysis. Both normal operation and hypothetical accident thermal transient conditions are evaluated. The BR-100 cask has a concrete layer than contains free water. During a hypothetical accident, the free water vaporizes and flows from the cask, removing a significant amount of thermal transient energy. The BR-100 transportation package meets the thermal requirements of 10CFR71. It additionally offers substantial margins to established material temperature limits.

  4. 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.

  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. 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.

  7. 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.

  8. Evaluation of the /sup 252/Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

    SciTech Connect

    Mihalczo, J.T.

    1987-11-15

    The /sup 252/Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt LWR fuel submerged in fuel storage pools, fully loaded and as they are being loaded. The motivation for this evaluation was that measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This in turn could lead to more cost-effective cask designs. Evaluation of the method for this application was based on (1) experiments already completed at a critical experiments facility using arrays of PWR fuel pins typical of the size of storage cask configurations, (2) the existence of neutron detectors that can function in shipping cask environments, and (3) the ability to construct ionization chambers containing /sup 252/Cf of adequate intensity for these measurements. These three considerations are discussed.

  9. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... International. SAR Title: Final Safety Analysis Report for the HI-STAR 100 Cask System. Docket Number: 72-1008. Certificate Expiration Date: October 4, 2019. Model Number: HI-STAR 100. Certificate Number: 1014....

  10. 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.

  11. 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 Storage...

  12. 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.

  13. 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.

  14. 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

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

    SciTech Connect

    Tsoukalas, Lefteri H.

    2016-03-01

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

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Transportation fuels: Desulfurizing diesel

    NASA Astrophysics Data System (ADS)

    Lamonier, Carole

    2017-02-01

    Transportation fuels such as diesel contain organosulfur molecules that, when combusted, form sulfur oxides that are toxic and poison vehicles' catalytic convertors. Now, a method is demonstrated that can reduce the sulfur concentration of diesel fuel to very low levels at low temperatures and pressures.

  1. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-18

    ... of the CoC Technical Specifications (TS) to clarify that the maximum design basis earthquake...) in appendix A of the CoC to clarify that the maximum design basis earthquake accelerations of 0.37g... earthquake accelerations of 0.37g in the horizontal direction (without cask sliding) and 0.25g in the...

  2. Test Plan for Cask Identification Detector

    SciTech Connect

    Rauch, Eric Benton

    2016-09-29

    This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench-scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

  3. 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.

  4. Apollo 12 Mission image - Modular Equipment Stowage Assemble (MESA) and the Fuel Cask on the Lunar Module (LM)

    NASA Image and Video Library

    1969-11-19

    AS12-48-7034 (19 Nov. 1969) --- A close-up view of a portion of quadrant II of the descent stage of the Apollo 12 Lunar Module (LM), photographed during the Apollo 12 extravehicular activity (EVA). At lower left is the LM's Y footpad. The empty Radioisotope Thermoelectric Generator (RTG) fuel cask is at upper right. The fuel capsule has already been removed and placed in the RTG. The RTG furnishes power for the Apollo Lunar Surface Experiments Package (ALSEP) which the Apollo 12 astronauts deployed on the moon. The LM's descent engine skirt is in the center background. The rod-like object protruding out from under the footpad is a lunar surface sensing probe. Astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) in lunar orbit while astronauts Charles Conrad Jr., commander; and Alan L. Bean, lunar module pilot, descended in the LM to explore the moon.

  5. Fuels for transportation.

    PubMed

    Fredholm, Bertil B; Nordén, Bengt

    2010-01-01

    There is a need to reduce the amount of fossil energy used for transport, both because of the easily available fossil fuel is becoming sparser and because of climate concerns. In this article, the concept of "peak oil" is briefly presented. Second, a practical approach to reduction of fossil fuel use for transport elaborated by two British commissions is presented. A key feature is the introduction of electric cars. This raises the third issue covered in this article: namely, how battery technology is going to meet the increasing needs posed by the transport sector.

  6. 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.

  7. 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.

  8. Spent fuel and HLW transportation the French experience

    SciTech Connect

    Giraud, J.P.; Charles, J.L.

    1995-12-31

    With 53 nuclear power plants in operation at EDF and a fuel cycle with recycling policy of the valuable materials, COGEMA is faced with the transport of a wide range of radioactive materials. In this framework, the transport activity is a key link in closing the fuel cycle. COGEMA has developed a comprehensive Transport Organization System dealing with all the sectors of the fuel cycle. The paper will describe the status of transportation of spent fuel and HLW in France and the experience gathered. The Transport Organization System clearly defines the role of all actors where COGEMA, acting as the general coordinator, specifies the tasks to be performed and brings technical and commercial support to its various subcontractors: TRANSNUCLEAIRE, specialized in casks engineering and transport operations, supplies packaging and performs transport operations, LEMARECHAL and CELESTIN operate transport by truck in the Vicinity of the nuclear sites while French Railways are in charge of spent fuel transport by train. HLW issued from the French nuclear program is stored for 30 years in an intermediate storage installation located at the La Hague reprocessing plant. Ultimately, these canisters will be transported to the disposal site. COGEMA has set up a comprehensive transport organization covering all operational aspects including adapted procedures, maintenance programs and personnel qualification.

  9. 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

  10. Barge shipment and reactor handling of a castor V/21 cask

    SciTech Connect

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

    1988-01-01

    The results of this study consist of a complete handling time/dose assessment for barge transport and reactor loading of a Castor V/21 storage cask. Observations are based on the barge transport and spent-fuel loading of storage casks at the Surry, Virginia, nuclear power plant during 1987. The minimum time required to perform all storage cask-handling activities from ship off-loading through placement of the loaded cask in at-reactor storage was 43.8 h. The addition of delays (due to backshifts not worked, etc.) resulted in a total turnaround time for the operation of {approx}6 days (24 h/day). Total labor requirement was 136 person-hours. Occupational dose for these activities totaled 416 person-mrem of exposure was due to background dose, representing {approx}40% of total dose. The highest dose-producing activity consisted of those steps involved with draining the loaded storage cask (i.e., installing drain pipe and pumping water from cask). This activity resulted in 75 person-mrem of exposure. Lid installation and vacuum drying of the cavity resulted in 56 person-mrem of exposure. The actual loading of spent-fuel assemblies into the storage cask was the third highest dose-contributing activity, resulting in 38 person-mrem of exposure.

  11. 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.

  12. 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.

  13. Estimation of the Transportation Risks for the Spent Fuel in Korea for Various Transportation Scenarios

    SciTech Connect

    Jongtae, Jeong; Cho, D.K.; Choi, H.J.; Choi, J.W.

    2008-07-01

    According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for the spent fuels in Korea. Also, we estimated and compared the transportation risks for these four transportation scenarios. From the results of this study, we found that these four transportation scenarios for spent fuels have a very low radiological risk activity with a manageable safety and health consequences. The results of this study can be used as basic data for the development of safe and economical logistics for a transportation of the spent fuels in Korea by considering the transportation costs for the four scenarios which will be needed in the near future. (authors)

  14. Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

    SciTech Connect

    Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

    2003-02-25

    Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research.

  15. 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.

  16. Transportation Fuels Markets, PADD 5

    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.

  17. Human machine interface to manually drive rhombic like vehicles such as transport casks in ITER

    SciTech Connect

    Lopes, Pedro; Vale, Alberto; Ventura, Rodrigo

    2015-07-01

    The Cask and Plug Remote Handling System (CPRHS) and the respective Cask Transfer System (CTS) are designed to transport activated components between the reactor and the hot cell buildings of ITER during maintenance operations. In nominal operation, the CPRHS/CTS shall operate autonomously under human supervision. However, in some unexpected situations, the automatic mode must be overridden and the vehicle must be remotely guided by a human operator due to the harsh conditions of the environment. The CPRHS/CTS is a rhombic-like vehicle with two independent steerable and drivable wheels along its longitudinal axis, giving it omni-directional capabilities. During manual guidance, the human operator has to deal with four degrees of freedom, namely the orientations and speeds of two wheels. This work proposes a Human Machine Interface (HMI) to manage the degrees of freedom and to remotely guide the CPRHS/CTS in ITER taking the most advantages of rhombic like capabilities. Previous work was done to drive each wheel independently, i.e., control the orientation and speed of each wheel independently. The results have shown that the proposed solution is inefficient. The attention of the human operator becomes focused in a single wheel. In addition, the proposed solution cannot assure that the commands accomplish the physical constrains of the vehicle, resulting in slippage or even in clashes. This work proposes a solution that consists in the control of the vehicle looking at the position of its center of mass and its heading in the world frame. The solution is implemented using a rotational disk to control the vehicle heading and a common analogue joystick to control the vector speed of the center of the mass of the vehicle. The number of degrees of freedom reduces to three, i.e., two angles (vehicle heading and the orientation of the vector speed) and a scalar (the magnitude of the speed vector). This is possible using a kinematic model based on the vehicle Instantaneous

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. Life cycle cost report of VHLW cask

    SciTech Connect

    1995-06-01

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

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

    SciTech Connect

    Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.; Mironov, B.S.; Kaplenkov, V.N.; Seredenko, A.V.; Saranchin, V.K.; Shulgin, A.S.; Haire, M.J.; Forsberg, C.W.

    2007-07-01

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

  7. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved...

  8. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved spent... Standardized NUHOMS® Horizontal Modular Storage System for Irradiated Nuclear Fuel. Docket Number: 72-1004...

  9. 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.

  10. Cask system design guidance for robotic handling

    SciTech Connect

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

    1990-10-01

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

  11. 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.

  12. 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.

  13. 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).

  14. Spent nuclear fuel system dynamic stability under normal conditions of transportation

    DOE PAGES

    Jiang, Hao; Wang, Jy-An John

    2016-10-14

    In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

  15. Spent nuclear fuel system dynamic stability under normal conditions of transportation

    SciTech Connect

    Jiang, Hao; Wang, Jy-An John

    2016-10-14

    In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.

  16. 76 FR 70374 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-14

    ... surface contamination limits; TS 4.1.1 to add various boron-10 areal densities for use with Pressurized Water Reactor and Boiling Water Reactor baskets and to replace the fuel tube orthogonal pitch with...

  17. 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

    ... ambient temperature from - 20 F to -21 F, expand the authorized contents of the NUHOMS HD System to include pressurized water reactor fuel assemblies with control components, reduce the minimum initial... neutron absorbers with integral aluminum cladding, delete use of nitrogen for draining the water from...

  18. 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

    ... water removal operations, and make corresponding changes to the technical specifications. DATES... minimum off-normal ambient temperature from -20 F to -21 F, expand the authorized contents of the NUHOMS HD System to include pressurized water reactor fuel assemblies with control components, reduce...

  19. 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

    ... contents; reduce the minimum off-normal ambient temperature from -20 F to -21 F; expand the authorized contents of the NUHOMS HD System to include pressurized water reactor fuel assemblies with control... requirements for neutron absorber tests; delete use of nitrogen for draining the water from the dry...

  20. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 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... has been determined by the NRC. The application must be accompanied by a safety analysis report (SAR...

  1. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-14

    ... nuclear fuel at civilian nuclear power reactor sites, with the objective of establishing one or more technologies that the Commission may, by rule, approve for use at the sites of civilian nuclear power reactors... site of any civilian nuclear power reactor.'' To implement this mandate, the Commission approved dry...

  2. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 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... has been determined by the NRC. The application must be accompanied by a safety analysis report (SAR...

  3. Use of burnup credit for transportation and storage

    SciTech Connect

    Sanders, T.L.; Ewing, R.I. ); Lake, W.H. )

    1991-01-01

    Burnup credit is the application of the effects of fuel burnup to nuclear criticality design. When burnup credit is considered in the design of storage facilities and transportation casks for spent fuel, the objectives are to reduce the requirements for storage space and to increase the payload of casks with acceptable nuclear criticality safety margins. The spent-fuel carrying capacities of previous-generation transport casks have been limited primarily by requirements to remove heat and/or to provide shielding. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus a considerable weight margin is available to the designer for increasing the payload capacity. One method to achieve an increase in capacity is to reduce fuel assembly spacing. The amount of reduction in assembly spacing is limited by criticality and fuel support structural concerns. The optimum fuel assembly spacing provides the maximum cask loading within a basket that has adequate criticality control and sufficient structural integrity for regulatory accident scenarios. The incorporation of burnup credit in cask designs could result in considerable benefits in the transport of spent fuel. The acceptance of burnup credit for the design of transport casks depends on the resolution of system safety issues and the uncertainties that affect the determination of criticality safety margins. The remainder of this report will examine these issues and the integrated approach under way to resolve them. 20 refs., 2 figs.

  4. 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.

  5. 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.

  6. 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…

  7. 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…

  8. Evaluation of RAPID for a UNF cask benchmark problem

    NASA Astrophysics Data System (ADS)

    Mascolino, Valerio; Haghighat, Alireza; Roskoff, Nathan J.

    2017-09-01

    This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for the simulation of a used nuclear fuel (UNF) cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days). A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.

  9. 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)

  10. 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.

  11. Methods of producing transportation fuel

    DOEpatents

    Nair, Vijay [Katy, TX; Roes, Augustinus Wilhelmus Maria [Houston, TX; Cherrillo, Ralph Anthony [Houston, TX; Bauldreay, Joanna M [Chester, GB

    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.

  12. Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

    SciTech Connect

    Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

    2016-09-01

    Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on

  13. 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

  14. Renewable and sustainable transportation fuels

    USDA-ARS?s Scientific Manuscript database

    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...

  15. 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.

  16. 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).

  17. 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.

  18. SCALE/MAVRIC calculation of dose rates measured for a gamma radiation source in a thick-walled transport and storage cask of ductile cast iron with lead inserts

    NASA Astrophysics Data System (ADS)

    Baumgarten, Werner; Thiele, Holger; Ruprecht, Benjamin; Phlippen, Peter-W.; Schlömer, Luc

    2017-09-01

    Dose rate calculations are important for judging the shielding performance of transport casks for radioactive material. Therefore it is important to have reliable calculation tools. We report on measured and calculated dose rates near a thick-walled transport and storage cask of ductile cast iron with lead inserts and a Co-60 source inside. In a series of experiments the thickness of the inserts was varied, and measured dose rates near the cask were compared with SCALE/MAVRIC 6.1.3 and SCALE/MAVRIC 6.2 calculation results. Deviations from the measurements were found to be higher for increased lead thicknesses. Furthermore, it is shown how the shielding material density, air scattering and accounting for the floor influence the quality of the calculation.

  19. 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.

  20. Surface storage cask test summarization report

    SciTech Connect

    Weeks, D L

    1983-09-01

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

  1. 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.

  2. Scoping study of casks shipped from the MRS facility to various repository sites

    SciTech Connect

    Bucholz, J.A.

    1986-08-01

    The objective of this study was to determine the maximum number of specialized repository waste packages that could be shipped from the Monitored Retrievable Storage (MRS) facility in Pb-, Fe-, and U-shielded casks weighing 200,000 or 300,000 lbs. The study included 18 different waste packages designed for the Salt, Tuff, and Basalt repositories. Nine of these contained consolidated PWR fuel pins, and nine contained consolidated BWR fuel pins. Discrete ordinates calculations were performed to determine the neutron and gamma shield thicknesses that would ensure a dose rate of 10 millirem/hr, 10 ft from the centerline of the cask(s). Over 100 casks of particular interest have been identified, while preliminary design information is also given for 522 casks of potential interest. Relative to the 200,000-lb casks, 50 to 100% more fuel may be shipped in the larger 300,000-lb casks. Placing the spent fuel canisters in overpacks prior to shipment from the MRS will reduce the net payload by 30 to 50%. The highest-capacity cask/waste package combination studied corresponds to a 300,000-lb U-shielded cask containing 84 consolidated PWR fuel assemblies in 21 canisters, or 171 consolidated BWR fuel assemblies in 19 canisters. Criticality analyses have shown these high-capacity casks to be safely subcritical - even if all the canisters were loaded with unirradiated LWR fuel containing 3.4 wt % U-235.

  3. 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.

  4. 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.

  5. Cask crush pad analysis using detailed and simplified analysis methods

    SciTech Connect

    Uldrich, E.D.; Hawkes, B.D.

    1997-12-31

    A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped spent nuclear fuel shipping cask into a 44-ft. deep cask unloading pool at the Fluorinel and Storage Facility (FAST). This facility, located at the Idaho Chemical Processing Plant (ICPP) at the Idaho national Engineering and Environmental Laboratory (INEEL), is a US Department of Energy site. The basis for this study is an analysis by Uldrich and Hawkes. The purpose of this analysis was to evaluate various hypothetical cask drop orientations to ensure that the crush pad design was adequate and the cask deceleration at impact was less than 100 g. It is demonstrated herein that a large spent fuel shipping cask, when dropped onto a foam crush pad, can be analyzed by either hand methods or by sophisticated dynamic finite element analysis using computer codes such as ABAQUS. Results from the two methods are compared to evaluate accuracy of the simplified hand analysis approach.

  6. 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

  7. Alternate-fueled transport aircraft possibilities

    NASA Technical Reports Server (NTRS)

    Aiken, W. S.

    1977-01-01

    The paper is organized to describe: (1) NASA's cryogenically fueled aircraft program; (2) LH2 subsonic and supersonic transport design possibilities (3) the fuel system and ground side problems associated with LH2 distribution; (4) a comparison of LCH4 with LH2; (5) the design possibilities for LCH4 fueled aircraft; and (6) a summary of where NASA's cryogenically fueled programs are headed.

  8. Methods of making transportation fuel

    DOEpatents

    Roes, Augustinus Wilhelmus Maria [Houston, TX; Mo, Weijian [Sugar Land, TX; Muylle, Michel Serge Marie [Houston, TX; Mandema, Remco Hugo [Houston, TX; Nair, Vijay [Katy, TX

    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.

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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)

  15. Conceptual design for an on-site spent-fuel transfer system

    SciTech Connect

    Pennington, C.W.; Neider, T.J.; Guerra, G.V.; Margotta, K.V. )

    1991-09-01

    Transnuclear, Inc. has developed two conceptual designs for transferring spent fuel from a nuclear plant's fuel storage pool to large storage casks by means of a small transfer cask. The transfer of spent fuel would be performed external to the nuclear plant, rather than within the plant's fuel building or transport bay. The systems can also be used to transfer fuel between storage casks and transport casks. Both systems meet the design considerations for on-site spent-fuel transfer systems outlined in EPRI Report NP-6425, Design Considerations for On-Site Spent-Fuel Transfer Systems.'' One of the designs involves a dry transfer into the storage cask, while the other uses a wet transfer method. Both systems have been evaluated for economic and technical feasibility, licensability and practicality. This report describes each of the conceptual designs and outlines the operating procedures for each system. In addition, a risk and accident assessment, a logistical evaluation, and an economic analysis including target cost evaluations and system cost estimates are included. This report shows that it is technically feasible and that costs are economically justifiable to perform on-site spent-fuel transfers using small transfer casks. However, the final design selection for the transfer system could depend on the individual utility's operational preferences and the unique facility features that already exist at each nuclear plant. 9 tabs.

  16. 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.

  17. 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)

  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. 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.

  20. 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...

  1. 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...

  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.; 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.

  3. 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).

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Improved fuel cell system for transportation applications

    SciTech Connect

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

    1991-12-31

    This invention is comprised of 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.

  10. 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.

  11. 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.

  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. 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.

  14. 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.

  15. 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

    ... cavity water removal operations; and make corresponding changes to the technical specifications. In the... fuel assemblies as authorized contents; reduce the minimum off-normal ambient temperature from -20 F to -21 F; expand the authorized contents of the NUHOMS HD System to include pressurized water...

  16. 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.

  17. Transportation Fuels Markets, Midwest and Rocky Mountain

    EIA Publications

    2017-01-01

    A new study commissioned by the U.S. Energy Information Administration (EIA), finds that changes in North American energy markets over the past decade have strengthened the supply of transportation fuels including motor gasoline, distillates, and jet fuel in the Midwest and Rocky Mountain regions.

  18. 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.

  19. Heated transportable fuel cell cartridges

    DOEpatents

    Lance, Joseph R.; Spurrier, Francis R.

    1985-01-01

    A fuel cell stack protective system is made where a plurality of fuel cells, each containing liquid electrolyte subject to crystallization, is enclosed by a containing vessel, and where at least one electric heater is placed in the containing vessel and is capable of preventing electrolyte crystallization.

  20. 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)...

  1. 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.

  2. 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...

  3. Assessment of the safety of spent fuel transportation in urban environs

    SciTech Connect

    Sandoval, R.P.; Weber, J.P.; Levine, H.S.; Romig, A.D.; Johnson, J.D.; Luna, R.E.; Newton, G.J.; Wong, B.A.; Marshall, R.W. Jr.; Alvarez, J.L.

    1983-06-01

    The results of a program to provide an experimental data base for estimating the radiological consequences from a hypothetical sabotage attack on a light-water-reactor spent fuel shipping cask in a densely populated area are presented. The results of subscale and full-scale experiments in conjunction with an analytical modeling study are described. The experimental data were used as input to a reactor-safety consequence model to predict radiological health consequences resulting from a hypothetical sabotage attack on a spent-fuel shipping cask in the Manhattan borough of New York City. The results of these calculations are presented.

  4. Response of spent LWR fuel to extreme environments

    SciTech Connect

    Sandoval, R.P.; Burian, R.J.; Kok, K.D.; DiSalvo, R.; Balmert, M.E.; Freeman-Kelly, R.; Fentiman, A.W.

    1986-01-01

    The research reported in this paper addresses the radiological source term which could arise when irradiated fuel in transport from a commercial light water reactor is exposed to the extreme environments postulated for some transportation accidents, specifically those involving a fire. The release of spent fuel radionuclides to the environment requires a breach of both the cask and the fuel rod cladding. Past research has given significant emphasis to evaluating the response of the shipping cask to mechanical and/or thermal loads from hypothetical accidents. Less consideration has been given to evaluating the response of the fuel rods to these environments. In this paper, the response of the fuel rods to an extreme thermal event was experimentally evaluated and the quantity of solid fuel material that could be released from the fuel rods to the cask cavity was estimated. Briefly, the objectives of this study were as follows: (1) Identify those conditions within a transportation cask which might produce fuel-rod cladding failure, emphasizing conditions associated with fires, and (2) Determine by experiment and analysis the nature of the source term so produced. The release of radionuclides from coolant or deposits on the outer surfaces of the fuel assembly was not addressed in this study. 6 figs., 2 figs.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. MCO loading and cask loadout technical manual

    SciTech Connect

    PRAGA, A.N.

    1998-10-01

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

  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. 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.

  12. Status of the Virginia Power/DOE Cooperative Cask Testing/Demonstration Program: A video presentation

    SciTech Connect

    McKinnon, M.A.; Creer, J.M. ); Collantes, C.E. . Richland Operations Office)

    1990-01-01

    This paper is documentation of a video presentation and provides a brief summary of the Virginia power/US Department of Energy Cooperative Cask Testing/Demonstration Program. The program consists of two phases. The first phase has been completed and involved the unlicensed performance testing (heat transfer and shielding) of three metal spent fuel storage casks at the federally owned Idaho National Engineering Laboratory. The second phase is ongoing and consists of licensed demonstrations of standard casks from two different vendors and of one or two enhanced capacity casks. 6 refs., 1 tab.

  13. 77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-02

    ... criticality safety analyses of pressurized water reactor spent nuclear fuel (SNF) in transportation packages... Doc No: 2012-10618] NUCLEAR REGULATORY COMMISSION [NRC-2012-0100] Burnup Credit in the Criticality... the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.'' This...

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ...-Cooled Nuclear Power Reactor 1 Normal Conditions of Transport Environmental impact Heat (per irradiated.... per truck; 100 tons per cask per rail car. Traffic density: Truck Less than 1 per day. Rail Less...

  19. Alternative transport fuels from natural gas

    SciTech Connect

    Moreno, R. Jr.; Bailey, D.G.F.

    1989-01-01

    This paper examines the economics of using natural gas as an alternative fuel in transport vehicles including passenger cars, taxis, buses, and trucks. It compares the cost of using conventional fuels (gasoline and diesel) in these vehicles with that of retrofitting the vehicles and using natural gas-based fuels. These fuels include compressed natural gas (CNG), liquefied petroleum gas (LPG), methanol, and synthetic gasoline and diesel. The analysis includes various scenarios to establish break-even these competing fuels under different assumptions. The study concludes that slow-fill (also known as trickle-fill) CNG is the most economic of the natural gas based fuels. The vehicles most suited for the economic use of these fuels are captive vehicle fleets with relatively high mileage and a restricted range of operation, such as buses and taxis. However, to make the transition economically viable, oil prices would have to increase significantly from their present level and remain there long enough to make the investment in the transition worth while. At present levels of crude oil prices (about US $18 per barrel), little opportunity exists for the economic substitution of natural gas based fuels for gasoline and diesel. These conclusions are based on the economic factors considered. Environmental factors, which are both difficult to measure and of increasing importance, could, however, mitigate these results.

  20. Hydrogen-fueled polymer electrolyte fuel cell systems for transportation.

    SciTech Connect

    Ahluwalia, R.; Doss, E.D.; Kumar, R.

    1998-10-19

    The performance of a polymer electrolyte fuel cell (PEFC) system that is fueled directly by hydrogen has been evaluated for transportation vehicles. The performance was simulated using a systems analysis code and a vehicle analysis code. The results indicate that, at the design point for a 50-kW PEFC system, the system efficiency is above 50%. The efficiency improves at partial load and approaches 60% at 40% load, as the fuel cell operating point moves to lower current densities on the voltage-current characteristic curve. At much lower loads, the system efficiency drops because of the deterioration in the performance of the compressor, expander, and, eventually, the fuel cell. The results also indicate that the PEFC system can start rapidly from ambient temperatures. Depending on the specific weight of the fuel cell (1.6 kg/kW in this case), the system takes up to 180s to reach its design operating conditions. The PEFC system has been evaluated for three mid-size vehicles: the 1995 Chrysler Sedan, the near-term Ford AIV (Aluminum Intensive Vehicle) Sable, and the future P2000 vehicle. The results show that the PEFC system can meet the demands of the Federal Urban Driving Schedule and the Highway driving cycles, for both warm and cold start-up conditions. The results also indicate that the P2000 vehicle can meet the fuel economy goal of 80 miles per gallon of gasoline (equivalent).

  1. High-capacity, high-strength trailer designs for the GA-4/GA-9 Casks

    SciTech Connect

    Kissinger, J.A.; Rickard, N.D.; Taylor, C.; Zimmer, A.

    1991-01-01

    General Atomics (GA) is developing final designs for two dedicated legal-weight trailers to transport the GA-4 and GA-9 Spent-Fuel Casks. The basic designs for these high-capacity, high-strength trailers are essentially identical except for small modifications to account for the differences in cask geometry. We are designing both trailers to carry a 55,000 lb (24,900 kg) payload and to withstand a 2.5 g vertical design load. The GA-4 and GA-9 trailers are designed for significantly higher loads than are typical commercial semitrailers, which are designed to loads in the range of 1.7 to 2.0 g. To meet the federal gross vehicle weight limit for legal-weight trucks, GA has set a target design weight for the trailers of 9000 lb (4080 kg). This weight includes the personnel barrier, cask tiedowns, and impact limiter removal and storage system. Based on the preliminary trailer designs, the final design weight is expected to be very close to this target weight. 3 refs., 3 figs.

  2. Transportation Fuels and the Hydrogen Economy

    NASA Astrophysics Data System (ADS)

    Gabbard, Alex

    2004-11-01

    An energy analysis of transportation fuels is performed for comparing automobiles and fuels currently in the marketplace as real world benchmarks projected as "hydrogen economy" requirements. Comparisons are made for ideal case average energy values at Standard Temperature and Pressure (STP) at 20°C, 1 atmosphere with no loses. "Real world" benchmarks currently in the marketplace illuminate the challenges to be met if an equivalent "hydrogen economy" is to become reality. The idea of a "hydrogen economy" is that, at some time in the future, world energy needs will be supplied in part or totally from hydrogen; in part as compared to the current "petroleum economy" that is the source of most of the world's transportation fuels and only a portion of total energy use, or hydrogen as the source of all energy consumption.

  3. 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.

  4. 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.

  5. Feasibility Study of Hydrogen Fueled Supersonic Transport

    NASA Astrophysics Data System (ADS)

    Yuhara, Tatsunori; Rinoie, Kenichi

    Recently, environmental problems have been paid great attention in the world. According to the current report of ICAO, CO2 emissions of aircraft to the total emissions will increase at an average annual rate of 3˜4%. Furthermore, stable supply of oil is becoming difficult. Without exceptions, those are hitting aviation industries. Therefore, in the near future, the time to shift from kerosene to liquid hydrogen (LH2) as fuel of aircraft will come. An idea of supersonic transport (SST) fueled by LH2 as one of the next-generation SST has been proposed by JAXA. The purpose of this study is to compare the weight, aerodynamics and cost of hydrogen fueled SST with those of kerosene fueled SST and to discuss its feasibilities. A possibility of two-bubble fuselage configuration is also evaluated.

  6. 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.

  7. 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

  8. Reactivity measurements on an experimental assembly of 4. 31 wt % sup 235 U enriched UO sub 2 fuel rods arranged in a shipping cask geometry

    SciTech Connect

    Bierman, S.R.

    1989-10-01

    A research program was initiated for the US Department of Energy (DOE) Sandia National Laboratory Transportation Systems Development Department in 1982 to provide benchmark type experimental criticality data in support of the design and safe operations of nuclear fuel transportation systems. The overall objective of the program is to identify and provide the experimental data needed to form a consistent, firm, and complete data base for verifying calculational models used in the criticality analyses of nuclear transport and related systems. A report, PNL-6205, issued in June 1988 (Bierman 1988) covered measurement results obtained from a series of experimental assemblies (TIC-1, 2, 3 and 4) involving neutron flux traps. The results obtained on a fifth experimental assembly (TIC-5), modeled after a calculational problem of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) Working Group, are covered in this report. 10 refs., 10 figs., 7 tabs.

  9. Transport of spent fuel from reactors to DOE storage/disposal facilities - a parametric study

    SciTech Connect

    Johnson, E.R.; Saverot, P.M.

    1997-09-01

    The amount of spent fuel from US commercial nuclear power plants that will be shipped to US DOE disposal facilities is expected to peak out at about 3000 MTU. A number of concerns including safety, emergency response, and traffic impacts, have been expressed regarding the large number of shipments that will be required to move this spent fuel. This article develops parametrically the numbers of shipments required to move 3000 MTU/year in the form of spent fuel under a variety of cask capacities and assumptions of fuel characteristics as well as the number of casks needed. In addition incremental traffic impact on roads and rail lines under a variety of conditions is also analyzed. 3 refs., 3 tabs.

  10. 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.

  11. 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.

  12. Environmental economics of lignin derived transport fuels.

    PubMed

    Obydenkova, Svetlana V; Kouris, Panos D; Hensen, Emiel J M; Heeres, Hero J; Boot, Michael D

    2017-11-01

    This paper explores the environmental and economic aspects of fast pyrolytic conversion of lignin, obtained from 2G ethanol plants, to transport fuels for both the marine and automotive markets. Various scenarios are explored, pertaining to aggregation of lignin from several sites, alternative energy carries to replace lignin, transport modalities, and allocation methodology. The results highlight two critical factors that ultimately determine the economic and/or environmental fuel viability. The first factor, the logistics scheme, exhibited the disadvantage of the centralized approach, owing to prohibitively expensive transportation costs of the low energy-dense lignin. Life cycle analysis (LCA) displayed the second critical factor related to alternative energy carrier selection. Natural gas (NG) chosen over additional biomass boosts well-to-wheel greenhouse gas emissions (WTW GHG) to a level incompatible with the reduction targets set by the U.S. renewable fuel standard (RFS). Adversely, the process' economics revealed higher profits vs. fossil energy carrier. Copyright © 2017 The Author(s). Published by Elsevier Ltd.. All rights reserved.

  13. 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.

  14. Final report on shipping-cask sabotage source-term investigation

    SciTech Connect

    Schmidt, E W; Walters, M A; Trott, B D; Gieseke, J A

    1982-10-01

    A need existed to estimate the source term resulting from a sabotage attack on a spent nuclear fuel shipping cask. An experimental program sponsored by the US NRC and conducted at Battelle's Columbus Laboratories was designed to meet that need. In the program a precision shaped charge was fired through a subscale model cask loaded with segments of spent PWR fuel rods and the radioactive material released was analyzed. This report describes these experiments and presents their results.

  15. 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

  16. 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.

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

    SciTech Connect

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

    2016-12-30

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

  18. 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.

  19. 327 SNF fuel return to K-Basin quality process plan

    SciTech Connect

    Ham, J.E.

    1998-09-22

    The B and W Hanford Company`s (BWHC) 327 Facility, in the 300 Area of the Hanford Site, contains Spent Nuclear Fuel (SNF) single fuel element canisters (SFEC) and fuel remnant canisters (FRC) which are to be returned to K-Basin. Seven shipments of up to six fuel canisters will be loaded into the CNS 1-13G Cask and transported to 105-KE.

  20. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect

    Durbin, Samuel G.; Lindgren, Eric R.

    2016-09-30

    The thermal performance of commercial nuclear spent fuel dry storage casks is 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 also by increasing the internal convection through greater canister helium pressure. These same 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 belowground 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 aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration.

  1. Polymer electrolyte fuel cells for transportation applications

    NASA Astrophysics Data System (ADS)

    Springer, T. E.; Wilson, M. S.; Garzon, F. H.; Zawodzinski, T. A.; Gottesfeld, S.

    The application of the polymer electrolyte fuel cell (PEFC) as a primary power source in electric vehicles has received increasing attention during the last few years. This increased attention has been fueled by a combination of significant technical advances in this field and by the initiation of some projects for the demonstration of a complete, PEFC-based power system in a bus or in a passenger car. Such demonstration projects reflect an increased faith of industry in the potential of this technology for transportation applications, or, at least, in the need for a detailed evaluation of this potential. Nevertheless, large scale transportation applications of PEFC's require a continued concerted effort of research on catalysis, materials and components, combined with the engineering efforts addressing the complete power system. This is required to achieve cost effective, highly performing PEFC stack and power system. We describe in this contribution some recent results of work performed within the Core Research PEFC Program at Los Alamos National Laboratory, which has addressed transportation applications of PEFC's.

  2. 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.

  3. Sustainable fuel for the transportation sector.

    PubMed

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

    2007-03-20

    A hybrid hydrogen-carbon (H(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(2) and CO(2) recycled from the H(2)-CO to liquid conversion reactor. Modeling of this biomass to liquids process has identified several major advantages of the H(2)CAR 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 approximately 30% of the United States transportation fuel from the annual biomass of 1.366 billion tons, the H(2)CAR process shows the potential to supply the entire United States transportation sector from that quantity of biomass. (iii) The synthesized liquid provides H(2) storage in an open loop system. (iv) Reduction to practice of the H(2)CAR route has the potential to provide the transportation sector for the foreseeable future, using the existing infrastructure. The rationale of using H(2) in the H(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(2)CAR is that there is no additional CO(2) release to the atmosphere due to the replacement of petroleum with coal, thus eliminating the need to sequester CO(2).

  4. 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

  5. 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}.

  6. 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

  7. 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.

  8. 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.

  9. 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).

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    SciTech Connect

    Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  17. 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.

  18. Transportation Fuels Markets, PADD 1 and PADD 3

    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.

  19. Transportation Fuels Markets, PADD 1 and PADD 3

    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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  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.; 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.

  5. 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.

  6. 76 FR 67287 - Alternative Fuel Transportation Program; Alternative Fueled Vehicle Credit Program (Subpart F...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-31

    ... Fuel Transportation Program (AFTP or Program), by including EISA-specified electric drive vehicles and... Fueled Vehicle 3. Automobile 4. Dedicated Vehicle 5. Dual Fueled Vehicle 6. Electric Motor Vehicle and Electric-Hybrid Vehicle 7. Section 133--Identified Vehicles That Already Qualify as AFVs B. New Definitions...

  7. Cost reductions of fuel cells for transport applications: fuel processing options

    NASA Astrophysics Data System (ADS)

    Teagan, W. P.; Bentley, J.; Barnett, B.

    The highly favorable efficiency/environmental characteristics of fuel cell technologies have now been verified by virtue of recent and ongoing field experience. The key issue regarding the timing and extent of fuel cell commercialization is the ability to reduce costs to acceptable levels in both stationary and transport applications. It is increasingly recognized that the fuel processing subsystem can have a major impact on overall system costs, particularly as ongoing R&D efforts result in reduction of the basic cost structure of stacks which currently dominate system costs. The fuel processing subsystem for polymer electrolyte membrane fuel cell (PEMFC) technology, which is the focus of transport applications, includes the reformer, shift reactors, and means for CO reduction. In addition to low cost, transport applications require a fuel processor that is compact and can start rapidly. This paper describes the impact of factors such as fuel choice, operating temperature, material selection, catalyst requirements, and controls on the cost of fuel processing systems. There are fuel processor technology paths which manufacturing cost analyses indicate are consistent with fuel processor subsystem costs of under 150/kW in stationary applications and 30/kW in transport applications. As such, the costs of mature fuel processing subsystem technologies should be consistent with their use in commercially viable fuel cell systems in both application categories.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. Liquid-fueled SOFC power sources for transportation

    SciTech Connect

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

    1994-11-01

    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.

  13. 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

  14. 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...

  15. 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...

  16. 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.

  17. 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.

  18. Alternative transport fuels from natural gas. Technical paper

    SciTech Connect

    Moreno, R.; Bailey, D.G.F.

    1989-01-01

    This report examines the economics of using natural gas as an alternative fuel in transport vehicles including passenger cars, taxis, buses, and trucks. It compares the cost of using conventional fuels (gasoline and diesel) in these vehicles with that of retrofitting the vehicles and using natural gas-based fuels. These fuels include compressed natural gas (CNG), liquefied petroleum gas (LPG), methanol, and synthetic gasoline and diesel. The analysis includes various scenarios to establish break-even points between these competing fuels under different assumptions.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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...

  6. 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...

  7. Fuel cells for electric utility and transportation applications

    SciTech Connect

    Srinivasan, S.

    1980-01-01

    This review article presents: the current status and expected progress status of the fuel cell research and development programs in the USA, electrochemical problem areas, techno-economic assessments of fuel cells for electric and/or gas utilities and for transportation, and other candidate fuel cells and their applications. For electric and/or gas utility applications, the most likely candidates are phosphoric, molten carbonate, and solid electrolyte fuel cells. The first will be coupled with a reformer (to convert natural gas, petroleum-derived, or biomass fuels to hydrogen), while the second and third will be linked with a coal gasifier. A fuel cell/battery hybrid power source is an attractive option for electric vehicles with projected performance characteristics approaching those for internal combustion or diesel engine powered vehicles. For this application, with coal-derived methanol as the fuel, a fuel cell with an acid electrolyte (phosphoric, solid polymer electrolyte or super acid) is essential; with pure hydrogen (obtained by splitting of water using nuclear, solar or hydroelectric energy), alkaline fuel cells show promise. A fuel cell researcher's dream is the development of a high performance direct methanol-air fuel cell as a power plant for electric vehicles. For long or intermittent duty cycle load leveling, regenerative hydrogen-halogen fuel cells exhibit desirable characteristics.

  8. 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.

  9. 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".

  10. Sensitivity analysis for best-estimate thermal models of vertical dry cask storage systems

    DOE PAGES

    DeVoe, Remy R.; Robb, Kevin R.; Skutnik, Steven E.

    2017-07-08

    Loading requirements for dry cask storage of spent nuclear fuel are driven primarily by decay heat capacity limitations, which themselves are determined through recommended limits on peak cladding temperature within the cask. This study examines the relative sensitivity of peak material temperatures within the cask to parameters that influence both the stored fuel residual decay heat as well as heat removal mechanisms. Here, these parameters include the detailed reactor operating history parameters (e.g., soluble boron concentrations and the presence of burnable poisons) as well as factors that influence heat removal, including non-dominant processes (such as conduction from the fuel basketmore » to the canister and radiation within the canister) and ambient environmental conditions. By examining the factors that drive heat removal from the cask alongside well-understood factors that drive decay heat, it is therefore possible to make a contextual analysis of the most important parameters to evaluation of peak material temperatures within the cask.« less

  11. 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).

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  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. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

    SciTech Connect

    Bowman, S.M.

    1991-04-01

    To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs.

  7. 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.

  8. Application of filtered back projection to muon radiography for imaging dry storage casks

    DOE PAGES

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

    2017-10-22

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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

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

    ScienceCinema

    Lewis, Nate

    2016-07-12

    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.

  14. 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.

  15. PEM fuel cells for transportation and stationary power generation applications

    SciTech Connect

    Cleghorn, S.J.; Ren, X.; Springer, T.E.; Wilson, M.S.; Zawodzinski, C.; Zawodzinski, T.A. Jr.; Gottesfeld, S.

    1996-05-01

    We describe recent activities at LANL devoted to polymer electrolyte fuel cells in the contexts of stationary power generation and transportation applications. A low cost/high performance hydrogen or reformate/air stack technology is being developed based on ultralow Pt loadings and on non-machined, inexpensive elements for flow-fields and bipolar plates. On board methanol reforming is compared to the option of direct methanol fuel cells because of recent significant power density increases demonstrated in the latter.

  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. 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. 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.

  19. 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.

  20. 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.

  1. 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)

  2. Consideration of Fuel Requirements for Supersonic Transport Operation

    NASA Technical Reports Server (NTRS)

    Stickle, Joseph W.

    1965-01-01

    An analysis of the interaction of operational environment and aircraft characteristics of the supersonic transport (SST) in the areas of design-range and reserve-fuel requirements has been made. Design-range requirements are considered in relation to the effects of wind, temperature, flight-level assignment, and payload variation. An approach toward combining en route and holding reserve requirements while maintaining protection equivalent to that provided subsonic jet transport operations by the present civil air regulation en route plus holding reserves is given. This approach results in a savings in reserve fuel over that required by separate requirements.

  3. 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.

  4. 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.

  5. Environmental considerations of a fuel-flexible transportation system

    SciTech Connect

    Saricks, C.L.

    1988-01-01

    The United States Department of Energy Office of Policy, Planning, and Analysis in conjuction with the Office of Environmental Analysis is conducting a multi-faceted, multi-organization study of the prospects for and costs of implementing a fuel-flexible (multifuel) domestic transportation system by the year 2000. The analysis has three components: (1) forecasting the supply of candidate alternative fuels (chiefly methanol, compressed natural gas, and electricity) by about year 2000, and of propulsion systems to use them; (2) projecting vehicular sales, fuel distributing and dispensing, and vehicular service infrastructures needed to support a ''best case'' penetration of each of these fuels individually or in combination into the national automobile, light and medium truck, and bus populations; and (3) considering changes in the environmental effects of transportation activity that could result from the fleet penetrations implicit in these cases. This paper presents the currently-proposed objectives and scope of the last of these three areas of analysis, and reviews environmental issues already identified as potential concerns for a period of transition to non-petroleum fuel dominance or at least of significant displacement of petroleum use in domestic transportation. These issues include performance and durability of vehicular emission control systems, formaldehyde transients of alcohol fuels, use of catalysts with heavy-duty vehicles, synergistic effects of gasoline/alcohol blends, safety of gas compressor systems, reduction of emissions of oxides of nitrogen, control of environmental release pathways for methane and toxic gases or solids, and institutional and legal constraints in some locations on the use of non-petroleum fuels in transportation. 20 refs., 1 fig.

  6. 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.

  7. Fuel Containment Concepts - Transport Category Airplanes.

    DTIC Science & Technology

    1987-11-01

    inhihi ting adhesive prim.r i’- app 1 d )rlor to bondin1 and they receive, an additional coat at ter bonding. Dense core is provided for stabi I itv in...installation of heat reticulated foam or expanded metal foil have the advantage of being passive systems. They prevent excessive overpressures from...Applicability of Reticulated Foams for the Suppression of Fuel Tank Explosions," AGARD-CP-166, Aircraft Fire Safety, Rome, Italy, April 1975. 45. MacDonald, J

  8. Recommendations for cask features for robotic handling from the Advanced Handling Technology Project

    SciTech Connect

    Drotning, W.

    1991-02-01

    This report describes the current status and recent progress in the Advanced Handling Technology Project (AHTP) initiated to explore the use of advanced robotic systems and handling technologies to perform automated cask handling operations at radioactive waste handling facilities, and to provide guidance to cask designers on the impact of robotic handling on cask design. Current AHTP tasks have developed system mock-ups to investigate robotic manipulation of impact limiters and cask tiedowns. In addition, cask uprighting and transport, using computer control of a bridge crane and robot, were performed to demonstrate the high speed cask transport operation possible under computer control. All of the current AHTP tasks involving manipulation of impact limiters and tiedowns require robotic operations using a torque wrench. To perform these operations, a pneumatic torque wrench and control system were integrated into the tool suite and control architecture of the gantry robot. The use of captured fasteners is briefly discussed as an area where alternative cask design preferences have resulted from the influence of guidance for robotic handling vs traditional operations experience. Specific robotic handling experiences with these system mock-ups highlight a number of continually recurring design principles: (1) robotic handling feasibility is improved by mechanical designs which emphasize operation with limited dexterity in constrained workspaces; (2) clearances, tolerances, and chamfers must allow for operations under actual conditions with consideration for misalignment and imprecise fixturing; (3) successful robotic handling is enhanced by including design detail in representations for model-based control; (4) robotic handling and overall quality assurance are improved by designs which eliminate the use of loose, disassembled parts. 8 refs., 15 figs.

  9. 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

  10. Optimization of microfluidic fuel cells using transport principles.

    PubMed

    Lee, Jinkee; Lim, Keng Guan; Palmore, G Tayhas R; Tripathi, Anubhav

    2007-10-01

    Microfluidic fuel cells exploit the lack of convective mixing at low Reynolds number to eliminate the need for a physical membrane to separate the fuel from the oxidant. Slow transport of reactants in combination with high catalytic surface-to-volume ratios often inhibit the efficiency of a microfluidic fuel cell. The performance of microfluidic devices that rely on surface electrochemical reactions is controlled by the interplay between reaction kinetics and the rate of mass transfer to the reactive surfaces. This paper presents theoretical and experimental work to describe the role of flow rate, microchannel geometry, and location of electrodes within a microfluidic fuel cell on its performance. A transport model, based on the convective-diffusive flux of reactants, is developed that describes the optimal conditions for maximizing both the average current density and the percentage of fuel utilized. The results show that the performance can be improved when the design of the device includes electrodes smaller than a critical length. The results of this study advance current approaches to the design of microfluidic fuel cells and other electrochemically-coupled microfluidic devices.

  11. 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.

    2014-09-25

    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

  12. 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.

  13. 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.

  14. 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).

  15. 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.

  16. 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

  17. Mass transport mechanism in porous fuel cell electrodes

    NASA Technical Reports Server (NTRS)

    Jonsson, I.; Lindholm, I.

    1969-01-01

    Results of experiments on hydrogen-oxygen fuel cells show that higher current densities are obtained with cell anodes having a 100 micron thin active layer of porous nickel containing silver electrocatalyst. Increase in current density is attributed to a convective mass transport mechanism.

  18. Fuel cells for electric utility and transportation applications

    NASA Astrophysics Data System (ADS)

    Srinivasan, S.

    1980-04-01

    The status of fuel cell development is reviewed. For terrestrial electric utility applications, the most promising are phosphoric acid, molten carbonate and solid electrolyte fuel cells. The first will be coupled with a reformer (to convert natural gas, petroleum derived and biomass fuels to hydrogen) while the second and third with a coal gasifier. As ground transportation power sources, the promising systems are phosphoric (or alternate acid) and alkaline electrolyte fuel cells. In the first case, methanol is most attractive while in the second, it will be hydrogen stored as a compressed gas or as a hydride. A technoeconomic assessment of a 'Regenerative Hydrogen-Halogen Energy Storage System' demonstrates the prospects of its use for load leveling when coupled with nuclear, solar or wind power plants.

  19. 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.

  20. Regional analysis of renewable transportation fuels - production and consumption

    NASA Astrophysics Data System (ADS)

    Liu, Xiaoshuai

    The transportation sector contributes more than a quarter of total U.S. greenhouse gas emissions. Replacing fossil fuels with renewable fuels can be a key solution to mitigate GHG emissions from the transportation sector. Particularly, we have focused on land-based production of renewable fuels from landfills and brownfield in the southeastern region of the United States. These so call marginal lands require no direct land-use change to avoid environmental impact and, furthermore, have rendered opportunities for carbon trading and low-carbon intensity business. The resources potential and production capacity were derived using federal and state energy databases with the aid of GIS techniques. To maximize fuels production and land-use efficiency, a scheme of co-location renewable transportation fuels for production on landfills was conducted as a case study. Results of economic modeling analysis indicate that solar panel installed on landfill sites could generate a positive return within the project duration, but the biofuel production within the landfill facility is relatively uncertain, requiring proper sizing of the onsite processing facility, economic scale of production and available tax credits. From the consumers' perspective, a life-cycle cost analysis has been conducted to determine the economic and environmental implications of different transportation choices by consumers. Without tax credits, only the hybrid electric vehicles have lifetime total costs equivalent to a conventional vehicles differing by about 1 to 7%. With tax credits, electric and hybrid electric vehicles could be affordable and attain similar lifetime total costs as compared to conventional vehicles. The dissertation research has provided policy-makers and consumers a pathway of prioritizing investment on sustainable transportation systems with a balance of environmental benefits and economic feasibility.

  1. 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

  2. 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

  3. Packaging design criteria for the MCO cask

    SciTech Connect

    Clements, M.D.

    1996-01-01

    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.

  4. Packaging design criteria for the MCO cask

    SciTech Connect

    Edwards, W.S.

    1997-01-30

    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.

  5. Oil Price Uncertainty, Transport Fuel Demand and Public Health.

    PubMed

    He, Ling-Yun; Yang, Sheng; Chang, Dongfeng

    2017-03-01

    Based on the panel data of 306 cities in China from 2002 to 2012, this paper investigates China's road transport fuel (i.e., gasoline and diesel) demand system by using the Almost Ideal Demand System (AIDS) and the Quadratic AIDS (QUAIDS) models. The results indicate that own-priceelasticitiesfordifferentvehiclecategoriesrangefrom-1.215to-0.459(byAIDS)andfrom -1.399 to-0.369 (by QUAIDS). Then, this study estimates the air pollution emissions (CO, NOx and PM2.5) and public health damages from the road transport sector under different oil price shocks. Compared to the base year 2012, results show that a fuel price rise of 30% can avoid 1,147,270 tonnes of pollution emissions; besides, premature deaths and economic losses decrease by 16,149 cases and 13,817.953 million RMB yuan respectively; while based on the non-linear health effect model, the premature deaths and total economic losses decrease by 15,534 and 13,291.4 million RMB yuan respectively. Our study combines the fuel demand and health evaluation models and is the first attempt to address how oil price changes influence public health through the fuel demand system in China. Given its serious air pollution emission and substantial health damages, this paper provides important insights for policy makers in terms of persistent increasing in fuel consumption and the associated health and economic losses.

  6. Oil Price Uncertainty, Transport Fuel Demand and Public Health

    PubMed Central

    He, Ling-Yun; Yang, Sheng; Chang, Dongfeng

    2017-01-01

    Based on the panel data of 306 cities in China from 2002 to 2012, this paper investigates China’s road transport fuel (i.e., gasoline and diesel) demand system by using the Almost Ideal Demand System (AIDS) and the Quadratic AIDS (QUAIDS) models. The results indicate that own-price elasticities for different vehicle categories range from −1.215 to −0.459 (by AIDS) and from −1.399 to −0.369 (by QUAIDS). Then, this study estimates the air pollution emissions (CO, NOx and PM2.5) and public health damages from the road transport sector under different oil price shocks. Compared to the base year 2012, results show that a fuel price rise of 30% can avoid 1,147,270 tonnes of pollution emissions; besides, premature deaths and economic losses decrease by 16,149 cases and 13,817.953 million RMB yuan respectively; while based on the non-linear health effect model, the premature deaths and total economic losses decrease by 15,534 and 13,291.4 million RMB yuan respectively. Our study combines the fuel demand and health evaluation models and is the first attempt to address how oil price changes influence public health through the fuel demand system in China. Given its serious air pollution emission and substantial health damages, this paper provides important insights for policy makers in terms of persistent increasing in fuel consumption and the associated health and economic losses. PMID:28257076

  7. 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.

  8. Rationale for converting the US transportation system to methanol fuel

    SciTech Connect

    Larsen, R.P.; Santini, D.J.

    1986-01-01

    An alternative fuel eventually suitable for a large part of the transportation system should be introduced into the US market on a large scale by the mid-1990s. A necessarily long transition must be carefully planned to avoid rapidly increasing fuel prices; dwindling domestic oil supplies; and overdependence on the Mideast, a geographically concentrated and potentially unreliable fuel source. This paper summarizes the growing evidence that supports methanol as the preferred alternative fuel. This evidence is now sufficiently compelling to make the risk of starting a transition to this fuel lower than the risk of failing to do so. In spite of the short-term oil glut and because of long-run oil depletion trends, the US government should implement a smooth transition to methanol fuel. The benefits associated with making this transition are described, and estimates are given of the aggregate benefit to the US economy. Finally, arguments are presented for appropriate involvement of the federal government in the transition, given the possible repercussions if the lessons of history are not heeded.

  9. 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.

  10. Water footprint of U.S. transportation fuels.

    PubMed

    Scown, Corinne D; Horvath, Arpad; McKone, Thomas E

    2011-04-01

    In the modern global economy, water and energy are fundamentally connected. Water already plays a major role in electricity generation and, with biofuels and electricity poised to gain a significant share of the transportation fuel market, water will become significantly more important for transportation energy as well. This research provides insight into the potential changes in water use resulting from increased biofuel or electricity production for transportation energy, as well as the greenhouse gas and freshwater implications. It is shown that when characterizing the water impact of transportation energy, incorporating indirect water use and defensible allocation techniques have a major impact on the final results, with anywhere between an 82% increase and a 250% decrease in the water footprint if evaporative losses from hydroelectric power are excluded. The greenhouse gas impact results indicate that placing cellulosic biorefineries in areas where water must be supplied using alternative means, such as desalination, wastewater recycling, or importation can increase the fuel's total greenhouse gas footprint by up to 47%. The results also show that the production of ethanol and petroleum fuels burden already overpumped aquifers, whereas electricity production is far less dependent on groundwater.

  11. 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.

  12. 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.

  13. 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.

  14. Packaging design criteria for the MCO cask

    SciTech Connect

    Edwards, W.S.

    1996-04-29

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins (including possibly 700 additional elements from PUREX, N Reactor, and 327 Laboratory). The basin water, particularly in the K East Basin, contains significant quantities of dissolved nuclear isotopes and radioactive fuel corrosion particles. 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. In order to initiate K Basin cleanup on schedule, the two-year fuel-shipping campaign must begin by December 1997. 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.

  15. Fuel Economy Through Teamwork. Energy Savings in School Transportation Publication Series. 1. Pupil Transportation and Energy Conservation. 2. Purchasing for Fuel Economy. 3. Driving for Fuel Economy. 4. Operating for Fuel Economy. 5. The Science of Saving Fuel.

    ERIC Educational Resources Information Center

    BRI Systems, Inc., Phoenix, AZ.

    This publication series of five booklets presents a summary of tips for saving energy in pupil transportation. The first booklet offers guidelines and suggestions to assist school transportation administration in achieving better fuel economy and cost management goals. The second presents purchasing tips and shows ways to use benefit cost analysis…

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Analysis of fuel ethanol transportation activity and potential distribution constraints

    SciTech Connect

    Das, Sujit; Peterson, Bruce E.; Chin, Shih -Miao

    2010-01-01

    This paper provides an analysis of fuel ethanol transportation activity and potential distribution constraints if the total 36 billion gallons of renewable fuel use by 2022 is mandated by EPA under the Energy Independence and Security Act (EISA) of 2007. Ethanol transport by domestic truck, marine, and rail distribution systems from ethanol refineries to blending terminals is estimated using Oak Ridge National Laboratory s (ORNL s) North American Infrastructure Network Model. Most supply and demand data provided by EPA were geo-coded and using available commercial sources the transportation infrastructure network was updated. The percentage increases in ton-mile movements by rail, waterways, and highways in 2022 are estimated to be 2.8%, 0.6%, and 0.13%, respectively, compared to the corresponding 2005 total domestic flows by various modes. Overall, a significantly higher level of future ethanol demand would have minimal impacts on transportation infrastructure. However, there will be spatial impacts and a significant level of investment required because of a considerable increase in rail traffic from refineries to ethanol distribution terminals.

  7. Evaluation of Transportation Options for Intermediate Non-destructive Examinations

    SciTech Connect

    Case, Susan; Hoggard, Gary

    2013-11-01

    Idaho National Laboratory (INL) shipments of irradiated experiments from the Advanced Test Reactor (ATR) to the Hot Fuels Examination Facility (HFEF) have historically been accomplished using the General Electric Model 2000 (GE 2000) Type B shipping container. Battelle Energy Alliance (BEA) concerns regarding the future availability and leasing and handling costs associated with the GE 2000 cask have warranted an evaluation of alternative shipping options. One or more of these shipping options may be utilized to perform non-destructive examinations (NDE) such as neutron radiography and precision gamma scans of irradiated experiments at HFEF and then return the experiments to ATR for further irradiation, hereafter referred to as “intermediate NDE.” This evaluation includes transportation options for intermediate NDE using the GE 2000 cask, BEA Research Reactor (BRR) package, Dry Transfer Cubicle (DTC) insert, and the General Electric Model 100 (GE 100) cask. The GE 2000 cask is the only Type B shipping container currently in use for shipments of irradiated material (exceeding Type A quantities) from ATR to HFEF; therefore it is included as one of the four shipping options in this evaluation. Cost and schedule estimates are provided for performing neutron radiography and precision gamma scans of a five-capsule drop-in-type ATR experiment for each transportation option. All costs provided in this evaluation are rough order-of-magnitude costs based on input from knowledgeable vendor employees and individuals at INL facilities.

  8. COBRA-SFS modifications and cask model optimization

    SciTech Connect

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

    1989-01-01

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

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

    SciTech Connect

    LANGEVIN, A.S.

    1999-07-12

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

  10. 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.

  11. Modeling transport in polymer-electrolyte fuel cells.

    PubMed

    Weber, Adam Z; Newman, John

    2004-10-01

    In this review, we have examined the different models for polymer-electrolyte fuel cells operating with hydrogen. The major focus has been on transport of the various species within the fuel cell. The different regions of the fuel cell were examined, and their modeling methodologies and equations were elucidated. In particular, the 1-D fuel-cell sandwich was discussed thoroughly because it is the most important part of the fuel-cell assembly. Models that included other effects such as temperature gradients and transport in other directions besides through the fuel-cell sandwich were also discussed. Models were not directly compared to each other; instead they were broken down into their constitutive parts. The reason for this is that validation of the models is usually accomplished by comparison of simulation to experimental polarization data (e.g., Figure 3). However, other data can also be used such as the net flux of water through the membrane. In fitting these data, the models vary not only in their complexity and treatments but also in their number and kind of fitting parameters. This is one reason it is hard to justify one approach over another by just looking at the modeling results. In general, it seems reasonable that the more complex models, which are based on physical arguments and do not contain many fitting parameters, are perhaps closest to reality. Of course, this assumes that they fit the experimental data and observations. This last point has been overlooked in the validation of many models. For example, a model may fit the data very well for certain operating conditions, but if it does not at least predict the correct trend when one of those conditions is changed, then the model is shown to be valid only within a certain operating range. This review has highlighted the important effects that should be modeled. These include two-phase flow of liquid water and gas in the fuel-cell sandwich, a robust membrane model that accounts for the different

  12. Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads due to Aircraft Engine Crash

    NASA Astrophysics Data System (ADS)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters(1) and seismic tests subjected to strong earthquake motions(2). Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001(3)-(6). This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are

  13. 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

  14. 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.

  15. Investigating fuel-cell transport limitations using hydrogen limiting current

    DOE PAGES

    Spingler, Franz B.; Phillips, Adam; Schuler, Tobias; ...

    2017-03-09

    Reducing mass-transport losses in polymer-electrolyte fuel cells (PEFCs) is essential to increase their power density and reduce overall stack cost. At the same time, cost also motivates the reduction in expensive precious-metal catalysts, which results in higher local transport losses in the catalyst layers. Here, we use a hydrogen-pump limiting-current setup to explore the gas-phase transport losses through PEFC catalyst layers and various gas-diffusion and microporous layers. It is shown that the effective diffusivity in the gas-diffusion layers is a strong function of liquid saturation. Additionally, it is shown how the catalyst layer unexpectedly contributes significantly to the overall measuredmore » transport resistance. This is especially true for low catalyst loadings. It is also shown how the various losses can be separated into different mechanisms including diffusional processes and mass-dependent and independent ones, where the data suggests that a large part of the transport resistance in catalyst layers cannot be attributed to a gas-phase diffusional process. The technique is promising for deconvoluting transport losses in PEFCs.« less

  16. 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.

  17. 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...

  18. HSCT research defines weight, fuel issues. [High Speed Civil Transport

    SciTech Connect

    Ott, J.

    1988-03-01

    The most recent research aims of the final stage of the NASA Langley High Speed Civil Transport (HSCT) studies program involve the definition of obstables to the commercial development of aircraft employing kerosene-based fuels for Mach 2-3 cruise. One of the manufacturers participating in the HSCT program finds maximum potential for initial operations of a Mach 2-2.4 SST in the 2000-2010 period; another manufacturer has extended the speed range to Mach 5, on the basis of cryogenic fuels. Attention has been given to the economics of operations between Pacific basin city pairs, such as Los Angeles-Sydney and Los Angeles-Tokyo, as well as between New York-Paris and New York-Tokyo.

  19. Analytical performance of direct-hydrogen-fueled polymer electrolyte fuel cell (PEFC) systems for transportation applications.

    SciTech Connect

    Doss, E. D.

    1998-06-02

    The performance of a stand-alone polymer electrolyte fuel cell (PEFC) system directly fueled by hydrogen has been evaluated for transportation vehicles. The study was carried out using a systems analysis code and a vehicle analysis code. The systems code includes models for the various PEFC components and is applicable for steady-state and transient situations. At the design point the system efficiency is above 50% for a 50-kW system. The efficiency improves under partial load and approaches 60% at 40% load, as the fuel cell operating point moves to lower current densities on the V-I polarization curve. At much lower loads, the system efficiency drops because of the deterioration in the performance of the compressor, expander, and eventually the fuel cell. The system performance suffers at lower temperatures, as the V-I characteristic curve for the fuel cell shifts downward because of the increased ohmic losses. The results of the transient analysis indicate that the hydrogen-fueled PEFC system can start rather rapidly, within seconds from ambient conditions. However, the warm-up time constant to reach the design operating temperatures is about 180 s. It is important during this period for the coolant to bypass the system radiator until the coolant temperature approaches the design temperature for the fuel cell. The systems analysis code has been applied to two mid-size vehicles: the near-term Ford AIV Sable and the future P2000 vehicle. The results of this study show that the PEFC system in these vehicles can respond well to the demands of the FUDS and Highway driving cycles, with both warm and cold starting conditions. The results also show that the fuel-cell AIV Sable vehicle has impressive gains in fuel economy over that of the internal combustion engine vehicle. However, this vehicle will not be able to meet the PNGV goal of 80 mpg. On the other hand, the P2000 vehicle approaches this goal with variable efficiency of the compressor and expander. It is expected

  20. Analytical performance of direct-hydrogen fueled polymer electrolyte fuel cell (PEFC) systems for transportation applications

    SciTech Connect

    Doss, E.D.; Ahluwalia, R.; Kumar, R.

    1998-07-01

    The performance of a stand-alone polymer electrolyte fuel cell (PEFC) system directly fueled by hydrogen has been evaluated for transportation vehicles. The study was carried out using a systems analysis code and a vehicle analysis code. The systems code includes models for the various PEFC components and is applicable for steady-state and transient situations. At the design point, the system efficiency is above 50% for a 50-kW system. The efficiency improves under partial load and approaches 60% at 40% load, as the fuel cell operating point moves to lower current densities on the V-I polarization curve. At much lower loads, the system efficiency drops because of the deterioration in the performance of the compressor, expander, and eventually the fuel cell. The system performance suffers at lower temperatures, as the V-I characteristic curve for the fuel cell shifts downward because of the increased ohmic losses. The results of the transient analysis indicate that the hydrogen-fueled PEFC system can start rather rapidly, within seconds from ambient conditions. However, the warm-up time constant to reach the design operating temperatures is about 180 s. It is important during this period for the coolant to bypass the system radiator until the coolant temperature approaches the design temperature for the fuel cell. The systems analysis code has been applied to two mid-size vehicles: the near-term Ford AIV Sable and the future P2000 vehicle. The results of this study show that the PEFC system in these vehicles can respond well to the demands of the FUDS and Highway driving cycles, with both warm and cold starting conditions. The results also show that the fuel-cell AIV Sable vehicle has impressive gains in fuel economy over that of the internal combustion engine vehicle. However, this vehicle will not be able to meet the PNGV goal of 80 mpg. On the other hand, the P2000 vehicle approaches this goal with variable efficiency of the compressor and expander. It is

  1. Particle transport in pellet fueled JET (Jet European Torus) plasmas

    SciTech Connect

    Baylor, L.R.

    1990-01-01

    Pellet fueling experiments have been carried out on the Joint European Torus (JET) tokamak with a multi-pellet injector. The pellets are injected at speeds approaching 1400 m/s and penetrate deep into the JET plasma. Highly peaked electron density profiles are achieved when penetration of the pellets approaches or goes beyond the magnetic axis, and these peaked profiles persist for more than two seconds in ohmic discharges and over one second in ICRF heated discharges. In this dissertation, analysis of electron particle transport in multi-pellet fueled JET limiter plasmas under a variety of heating conditions is described. The analysis is carried out with a one and one-half dimensional radial particle transport code to model the experimental density evolution with various particle transport coefficients. These analyses are carried out in plasmas with ohmic heating, ICRF heating, and neural beam heating, in limiter configurations. Peaked density profile cases are generally characterized by diffusion coefficients with a central (r/a < 0.5) diffusivity {approximately}0.1 m{sup 2}/s that increases rapidly to {approximately}0.3 m{sup 2}/s at r/a = 0.6 and then increases out to the plasma edge as (r/a){sup 2}. These discharges can be satisfactorily modeled without any anomalous convective (pinch) flux. 79 refs., 60 figs.

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

    SciTech Connect

    Santi, Peter A; Browne, Michael C; Williams, Richard B; Parker, Robert F

    2009-01-01

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

  3. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  4. 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.

  5. 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.

  6. Actinide partitioning-transmutation program. V. Preconceptual designs and costs of partitioning facilities and shipping casks, Appendix 4. Final report

    SciTech Connect

    Not Available

    1980-06-01

    This Appendix contains cost estimate documents for the Fuels Fabrication Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contributing to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed. Shipping cask costs are provided.

  7. Numerical Estimation of the Spent Fuel Ratio

    SciTech Connect

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

    2016-01-01

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

  8. Energy analysis and break-even distance for transportation for biofuels in comparison to fossil fuels

    USDA-ARS?s Scientific Manuscript database

    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...

  9. Fuel Cells for Electric Utility and Transportation Applications

    SciTech Connect

    Srinivasan, S.

    1980-01-01

    This review encompasses the following topics: (1) historical, (2) types of fuel cells, (3) thermodynamic and electrode kinetic aspects of fuel cells, (4) overview of present status of fuel cell research and development, (5) electrocatalysis of fuel cell reactions, (6) fuel cell/battery hybrid vehicles, and (7) regenerative hydrogen-halogen fuel cells for energy storage. (WHK)

  10. Transport properties of C and O in UN fuels

    NASA Astrophysics Data System (ADS)

    Schuler, Thomas; Lopes, Denise Adorno; Claisse, Antoine; Olsson, Pär

    2017-03-01

    Uranium nitride fuel is considered for fast reactors (GEN-IV generation and space reactors) and for light water reactors as a high-density fuel option. Despite this large interest, there is a lack of information about its behavior for in-pile and out-of-pile conditions. From the present literature, it is known that C and O impurities have significant influence on the fuel performance. Here we perform a systematic study of these impurities in the UN matrix using electronic-structure calculations of solute-defect interactions and microscopic jump frequencies. These quantities were calculated in the DFT +U approximation combined with the occupation matrix control scheme, to avoid convergence to metastable states for the 5 f levels. The transport coefficients of the system were evaluated with the self-consistent mean-field theory. It is demonstrated that carbon and oxygen impurities have different diffusion properties in the UN matrix, with O atoms having a higher mobility, and C atoms showing a strong flux coupling anisotropy. The kinetic interplay between solutes and vacancies is expected to be the main cause for surface segregation, as incorporation energies show no strong thermodynamic segregation preference for (001) surfaces compared with the bulk.

  11. 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.

  12. Solid oxide fuel cells for transportation: A clean, efficient alternative for propulsion

    SciTech Connect

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

    1993-04-01

    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, as am the major R&D issues.

  13. Solid oxide fuel cells for transportation: A clean, efficient alternative for propulsion

    SciTech Connect

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

    1993-01-01

    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, as am the major R D issues.

  14. 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.

  15. 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.

  16. 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.

  17. Importance of OH(-) transport from cathodes in microbial fuel cells.

    PubMed

    Popat, Sudeep C; Ki, Dongwon; Rittmann, Bruce E; Torres, César I

    2012-06-01

    Cathodic limitation in microbial fuel cells (MFCs) is considered an important hurdle towards practical application as a bioenergy technology. The oxygen reduction reaction (ORR) needs to occur in MFCs under significantly different conditions compared to chemical fuel cells, including a neutral pH. The common reason cited for cathodic limitation is the difficulty in providing protons to the catalyst sites. Here, we show that it is not the availability of protons, but the transport of OH(-) from the catalyst layer to the bulk liquid that largely governs cathodic potential losses. OH(-) is a product of an ORR mechanism that has not been considered dominant before. The accumulation of OH(-) at the catalyst sites results in an increase in the local cathode pH, resulting in Nernstian concentration losses. For Pt-based gas-diffusion cathodes, using polarization curves developed in unbuffered and buffered solutions, we quantified this loss to be >0.3 V at a current density of 10 Am(-2) . We show that this loss can be partially overcome by replacing the Nafion binder used in the cathode catalyst layer with an anion-conducting binder and by providing additional buffer to the cathode catalyst directly in the form of CO(2) , which results in enhanced OH(-) transport. Our results provide a comprehensive analysis of cathodic limitations in MFCs and should allow researchers to develop and select materials for the construction of MFC cathodes and identify operational conditions that will help minimize Nernstian concentration losses due to pH gradients.

  18. 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.

  19. 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

  20. NREL - Advanced Vehicles and Fuels Basics - Center for Transportation Technologies and Systems 2010

    ScienceCinema

    None

    2016-07-12

    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