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Sample records for cask mediate sorting

  1. Molecular basis for SNX-BAR-mediated assembly of distinct endosomal sorting tubules

    DEFF Research Database (Denmark)

    van Weering, Jan R.T.; Sessions, Richard B.; Traer, Colin J.;

    2012-01-01

    Sorting nexins (SNXs) are regulators of endosomal sorting. For the SNX-BAR subgroup, a Bin/Amphiphysin/Rvs (BAR) domain is vital for formation/stabilization of tubular subdomains that mediate cargo recycling. Here, by analysing the in vitro membrane remodelling properties of all 12 human SNX-BARs......, we report that some, but not all, can elicit the formation of tubules with diameters that resemble sorting tubules observed in cells. We reveal that SNX-BARs display a restricted pattern of BAR domain-mediated dimerization, and by resolving a 2.8 Å structure of a SNX1-BAR domain homodimer, establish...... that dimerization is achieved in part through neutralization of charged residues in the hydrophobic BAR-dimerization interface. Membrane remodelling also requires functional amphipathic helices, predicted to be present in all SNX-BARs, and the formation of high order SNX-BAR oligomers through selective...

  2. Vacuolar Sorting Receptor-Mediated Trafficking of Soluble Vacuolar Proteins in Plant Cells

    Directory of Open Access Journals (Sweden)

    Hyangju Kang

    2014-08-01

    Full Text Available Vacuoles are one of the most prominent organelles in plant cells, and they play various important roles, such as degradation of waste materials, storage of ions and metabolites, and maintaining turgor. During the past two decades, numerous advances have been made in understanding how proteins are specifically delivered to the vacuole. One of the most crucial steps in this process is specific sorting of soluble vacuolar proteins. Vacuolar sorting receptors (VSRs, which are type I membrane proteins, are involved in the sorting and packaging of soluble vacuolar proteins into transport vesicles with the help of various accessory proteins. To date, large amounts of data have led to the development of two different models describing VSR-mediated vacuolar trafficking that are radically different in multiple ways, particularly regarding the location of cargo binding to, and release from, the VSR and the types of carriers utilized. In this review, we summarize current literature aimed at elucidating VSR-mediated vacuolar trafficking and compare the two models with respect to the sorting signals of vacuolar proteins, as well as the molecular machinery involved in VSR-mediated vacuolar trafficking and its action mechanisms.

  3. Distinct pathways mediate the sorting of tail-anchored proteins to the plastid outer envelope.

    Directory of Open Access Journals (Sweden)

    Preetinder K Dhanoa

    Full Text Available BACKGROUND: Tail-anchored (TA proteins are a distinct class of membrane proteins that are sorted post-translationally to various organelles and function in a number of important cellular processes, including redox reactions, vesicular trafficking and protein translocation. While the molecular targeting signals and pathways responsible for sorting TA proteins to their correct intracellular destinations in yeasts and mammals have begun to be characterized, relatively little is known about TA protein biogenesis in plant cells, especially for those sorted to the plastid outer envelope. METHODOLOGY/PRINCIPAL FINDINGS: Here we investigated the biogenesis of three plastid TA proteins, including the 33-kDa and 34-kDa GTPases of the translocon at the outer envelope of chloroplasts (Toc33 and Toc34 and a novel 9-kDa protein of unknown function that we define here as an outer envelope TA protein (OEP9. Using a combination of in vivo and in vitro assays we show that OEP9 utilizes a different sorting pathway than that used by Toc33 and Toc34. For instance, while all three TA proteins interact with the cytosolic OEP chaperone/receptor, AKR2A, the plastid targeting information within OEP9 is distinct from that within Toc33 and Toc34. Toc33 and Toc34 also appear to differ from OEP9 in that their insertion is dependent on themselves and the unique lipid composition of the plastid outer envelope. By contrast, the insertion of OEP9 into the plastid outer envelope occurs in a proteinaceous-dependent, but Toc33/34-independent manner and membrane lipids appear to serve primarily to facilitate normal thermodynamic integration of this TA protein. CONCLUSIONS/SIGNIFICANCE: Collectively, the results provide evidence in support of at least two sorting pathways for plastid TA outer envelope proteins and shed light on not only the complex diversity of pathways involved in the targeting and insertion of proteins into plastids, but also the molecular mechanisms that underlie

  4. Rab7 associates with early endosomes to mediate sorting and transport of Semliki forest virus to late endosomes.

    Directory of Open Access Journals (Sweden)

    2005-07-01

    Full Text Available Semliki forest virus (SFV is internalized by clathrin-mediated endocytosis, and transported via early endosomes to late endosomes and lysosomes. The intracellular pathway taken by individual fluorescently labeled SFV particles was followed using immunofluorescence in untransfected cells, and by video-enhanced, triple-color fluorescence microscopy in live cells transfected with GFP- and RFP-tagged Rab5, Rab7, Rab4, and Arf1. The viruses progressed from Rab5-positive early endosomes to a population of early endosomes (about 10% of total that contained both Rab5 and Rab7. SFV were sequestered in the Rab7 domains, and they were sorted away from the early endosomes when these domains detached as separate transport carriers devoid of Rab5, Rab4, EEA1, Arf1, and transferrin. The process was independent of Arf1 and the acidic pH in early endosomes. Nocodazole treatment showed that the release of transport carriers was assisted by microtubules. Expression of constitutively inactive Rab7T22N resulted in accumulation of SFV in early endosomes. We concluded that Rab7 is recruited to early endosomes, where it forms distinct domains that mediate cargo sorting as well as the formation of late-endosome-targeted transport vesicles.

  5. The overexpressed human 46-kDa mannose 6-phosphate receptor mediates endocytosis and sorting of β-glucuronidase

    International Nuclear Information System (INIS)

    The authors studied the function of the human small (46-kDa) mannose 6-phosphate receptor (SMPR) in transfected mouse L cells that do not express the larger insulin-like growth factor II/mannose 6-phosphate receptor. Cells overexpressing human SMPR were studied for enzyme binding to cell surface receptors, for binding to intracellular receptors in permeabilized cells, and for receptor-mediated endocytosis of recombinant human β-glucuronidase. Specific binding to human SMPR in permeabilized cells showed a pH optimum between pH 6.0 and pH 6.5. Binding was significant in the present of EDTA but was enhanced by added divalent cations. Up to 2.3% of the total functional receptor could be detected on the cell surface by enzyme binding. They present experiments showing that at very high levels of overexpression, and at pH 6.5, human SMPR mediated the endocytosis of β-glucuronidase. At pH 7.5, the rate of endocytosis was only 14% the rate seen at pH 6.5. Cells overexpressing human SMPR also showed reduced secretion of newly synthesized β-glucuronidase when compared to cells transfected with vector only, suggesting that overexpressed human SMPR can participate in sorting of newly synthesized β-glucuronidase and partially correct the sorting defect in mouse L cells that do not express the insulin-like growth factor II/mannose 6-phosphate receptor

  6. Sorting Out Sorts

    OpenAIRE

    Jonathan B. Berk

    1998-01-01

    In this paper we analyze the theoretical implications of sorting data into groups and then running asset pricing tests within each group. We show that the way this procedure is implemented introduces a severe bias in favor of rejecting the model under consideration. By simply picking enough groups to sort into even the true asset pricing model can be shown to have no explanatory power within each group.

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

    Energy Technology Data Exchange (ETDEWEB)

    Newman, John T.; Mendez, Nicholas [IP Systems, Inc., 2685 Industrial Lane, Broomfield, Colorado 80020 (United States)

    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)

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

    International Nuclear Information System (INIS)

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

  9. Molecular determinants that mediate the sorting of human ATG9A from the endoplasmic reticulum.

    Science.gov (United States)

    Staudt, Catherine; Gilis, Florentine; Boonen, Marielle; Jadot, Michel

    2016-09-01

    ATG9A is a multispanning membrane protein required for autophagosome formation. Under basal conditions, neosynthesized ATG9A proteins travel to the Golgi apparatus and cycle between the trans-Golgi network and endosomes. In the present work, we searched for molecular determinants involved in the subcellular trafficking of human ATG9A in HeLa cells using sequential deletions and point mutations. Deletion of amino acids L(340) to L(354) resulted in the retention of ATG9A in the endoplasmic reticulum. In addition, we found that substitution of the L(711)YM(713) sequence (located in the C-terminal region of ATG9A) by alanine residues severely impaired its transport through the Golgi apparatus. This defect could be corrected by oligomerization of the mutant protein with co-transfected wild-type ATG9A, suggesting that ATG9A oligomerization may help its sorting through biosynthetic compartments. Lastly, the study of the consequences of the LYM/AAA mutation on the intracellular trafficking of ATG9A highlighted that some newly synthesized ATG9A can bypass the Golgi apparatus to reach the plasma membrane. Taken together, these findings provide new insights into the intracellular pathways followed by ATG9A to reach different subcellular compartments, and into the intramolecular determinants that drive the sorting of this protein. PMID:27316455

  10. Actin-Sorting Nexin 27 (SNX27)-Retromer Complex Mediates Rapid Parathyroid Hormone Receptor Recycling.

    Science.gov (United States)

    McGarvey, Jennifer C; Xiao, Kunhong; Bowman, Shanna L; Mamonova, Tatyana; Zhang, Qiangmin; Bisello, Alessandro; Sneddon, W Bruce; Ardura, Juan A; Jean-Alphonse, Frederic; Vilardaga, Jean-Pierre; Puthenveedu, Manojkumar A; Friedman, Peter A

    2016-05-20

    The G protein-coupled parathyroid hormone receptor (PTHR) regulates mineral-ion homeostasis and bone remodeling. Upon parathyroid hormone (PTH) stimulation, the PTHR internalizes into early endosomes and subsequently traffics to the retromer complex, a sorting platform on early endosomes that promotes recycling of surface receptors. The C terminus of the PTHR contains a type I PDZ ligand that binds PDZ domain-containing proteins. Mass spectrometry identified sorting nexin 27 (SNX27) in isolated endosomes as a PTHR binding partner. PTH treatment enriched endosomal PTHR. SNX27 contains a PDZ domain and serves as a cargo selector for the retromer complex. VPS26, VPS29, and VPS35 retromer subunits were isolated with PTHR in endosomes from cells stimulated with PTH. Molecular dynamics and protein binding studies establish that PTHR and SNX27 interactions depend on the PDZ recognition motif in PTHR and the PDZ domain of SNX27. Depletion of either SNX27 or VPS35 or actin depolymerization decreased the rate of PTHR recycling following agonist stimulation. Mutating the PDZ ligand of PTHR abolished the interaction with SNX27 but did not affect the overall rate of recycling, suggesting that PTHR may directly engage the retromer complex. Coimmunoprecipitation and overlay experiments show that both intact and mutated PTHR bind retromer through the VPS26 protomer and sequentially assemble a ternary complex with PTHR and SNX27. SNX27-independent recycling may involve N-ethylmaleimide-sensitive factor, which binds both PDZ intact and mutant PTHRs. We conclude that PTHR recycles rapidly through at least two pathways, one involving the ASRT complex of actin, SNX27, and retromer and another possibly involving N-ethylmaleimide-sensitive factor. PMID:27008860

  11. Testing of Metal Cask and Concrete Cask

    International Nuclear Information System (INIS)

    In Japan, the first interim spent fuel storage facility (ISF) outside of nuclear power plant site in use of dual-purpose metal cask is being planned to start its commercial operation in 2012 in Mutsu city, Aomori prefecture. The CRIEPI (Central Research Institute of Electric Power Industry) has executed several study programs on demonstrative testing for interim storage of spent fuel, mainly related to metal cask and concrete cask storage technology to reflect in Japanese safety requirements for dry casks issued by NISA/METI (Nuclear and Industrial Safety Agency, Ministry of Economy and Trade Industry). On top of that, the Japan Nuclear Energy Safety Organization (JNES) has executed study programs on spent fuel integrity, etc. This paper introduces the summary of these research programs. (author)

  12. Sorting of the Alzheimer's Disease Amyloid Precursor Protein Mediated by the AP-4 Complex

    Energy Technology Data Exchange (ETDEWEB)

    Burgos, Patricia V.; Mardones, Gonzalo A.; Rojas, Adriana L.; daSilva, Luis L.P.; Prabhu, Yogikala; Hurley, James H.; Bonifacino, Juan S. (NIH)

    2010-08-12

    Adaptor protein 4 (AP-4) is the most recently discovered and least well-characterized member of the family of heterotetrameric adaptor protein (AP) complexes that mediate sorting of transmembrane cargo in post-Golgi compartments. Herein, we report the interaction of an YKFFE sequence from the cytosolic tail of the Alzheimer's disease amyloid precursor protein (APP) with the {micro}4 subunit of AP-4. Biochemical and X-ray crystallographic analyses reveal that the properties of the APP sequence and the location of the binding site on 4 are distinct from those of other signal-adaptor interactions. Disruption of the APP-AP-4 interaction decreases localization of APP to endosomes and enhances {gamma}-secretase-catalyzed cleavage of APP to the pathogenic amyloid-{beta} peptide. These findings demonstrate that APP and AP-4 engage in a distinct type of signal-adaptor interaction that mediates transport of APP from the trans-Golgi network (TGN) to endosomes, thereby reducing amyloidogenic processing of the protein.

  13. Receptor-mediated sorting of soluble vacuolar proteins: myths, facts, and a new model.

    Science.gov (United States)

    Robinson, David G; Neuhaus, Jean-Marc

    2016-08-01

    To prevent their being released to the cell exterior, acid hydrolases are recognized by receptors at some point in the secretory pathway and diverted towards the lytic compartment of the cell (lysosome or vacuole). In animal cells, the receptor is called the mannosyl 6-phosphate receptor (MPR) and it binds hydrolase ligands in the trans-Golgi network (TGN). These ligands are then sequestered into clathrin-coated vesicles (CCVs) because of motifs in the cytosolic tail of the MPR which interact first with monomeric adaptors (Golgi-localized, Gamma-ear-containing, ARF-binding proteins, GGAs) and then with tetrameric (adaptin) adaptor complexes. The CCVs then fuse with an early endosome, whose more acidic lumen causes the ligands to dissociate. The MPRs are then recycled back to the TGN via retromer-coated carriers. Plants have vacuolar sorting receptors (VSRs) which were originally identified in CCVs isolated from pea (Pisum sativum L.) cotyledons. It was therefore assumed that VSRs would have an analogous function in plants to MPRs in animals. Although this dogma has enjoyed wide support over the last 20 years there are many inconsistencies. Recently, results have been published which are quite contrary to it. It now emerges that VSRs and their ligands can interact very early in the secretory pathway, and dissociate in the TGN, which, in contrast to its mammalian counterpart, has a pH of 5.5. Multivesicular endosomes in plants lack proton pump complexes and consequently have an almost neutral internal pH, which discounts them as organelles of pH-dependent receptor-ligand dissociation. These data force a critical re-evaluation of the role of CCVs at the TGN, especially considering that vacuolar cargo ligands have never been identified in them. We propose that one population of TGN-derived CCVs participate in retrograde transport of VSRs from the TGN. We also present a new model to explain how secretory and vacuolar cargo proteins are effectively separated after

  14. A cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. A cask fleet operations study

    International Nuclear Information System (INIS)

    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

  16. Cask storing facility

    International Nuclear Information System (INIS)

    The present invention provides a facility suitable to keeping and storing of casks for transporting and storing spent fuels generated from power plants and radioactive wastes generated from spent fuel reprocessing plants. Namely, the casks are transported in and out by a portal crane when they are stored. The cask storage space is disposed underground and soils are used as a portion of shielding materials. Then, a portal crane gives less load on the storage building when it is used compared with a case of using an overhead traveling crane. Since the storage pits are disposed underground, the radiation released from the casks in lateral and downward directions can be shielded by the soils. If shielding lids are disposed on the upper portion of the cask storage pits, upward radiation released from the casks can be shielded. Accordingly, there is no need to ensure thickness of walls of the building and ceilings for shielding. As a result, construction cost for the building can be reduced. (I.S.)

  17. A DI-LEUCINE SORTING SIGNAL IN ZIP1 (SLC39A1) MEDIATES ENDOCYTOSIS OF THE PROTEIN

    Science.gov (United States)

    It has been demonstrated that the plasma membrane expression of ZIP1 was regulated by endocytic mechanisms. In zinc-replete conditions, the level of surface expressed ZIP1 was low due to the rapid internalization of ZIP1 from the cytoplasmic membrane. The aim of the present work was to identify sort...

  18. The CD3 gamma leucine-based receptor-sorting motif is required for efficient ligand-mediated TCR down-regulation

    DEFF Research Database (Denmark)

    von Essen, Marina; Menné, Charlotte; Nielsen, Bodil L;

    2002-01-01

    other pathway is dependent on protein kinase C (PKC)-mediated activation of the CD3 gamma di-leucine-based receptor-sorting motif. Previous studies have failed to demonstrate a connection between ligand- and PKC-induced TCR down-regulation. Thus, although an apparent paradox, the dogma has been that...... ligand- and PKC-induced TCR down-regulations are not interrelated. By analyses of a newly developed CD3 gamma-negative T cell variant, freshly isolated and PHA-activated PBMC, and a mouse T cell line, we challenged this dogma and demonstrate in this work that PKC activation and the CD3 gamma di...

  19. Cask development, testing, and licensing

    International Nuclear Information System (INIS)

    The NuPac 125-B Rail Cask was developed to provide a safe means of transporting the damaged core of Three Mile Island Unit 2 from the TMI site at Middletown, PA, to the Idaho National Engineering laboratory (INEL) at Idaho Falls, ID. The development of the NuPac 125-B Rail Cask posed two engineering and technical management challenges; Licensing Strategy - The NuPac 125-B Rail Cask represented the first irradiated fuel rail cask developed within the United States in the past decade, a decade characterized by changing nuclear regulations, and Accelerated Schedule - The TMI-2 defueling schedule demanded a cask development schedule one-third as long as normally required. These challenges governed the overall development and licensing process for the cask. First, a high degree of conservation was incorporated into the design to allow quick, simplified demonstrations of adequacy to regulatory staff. Second, redundant design techniques were employed in all areas of uncertainty. The testing program eliminated performance uncertainties and validated predictions and predictive models. Drop tests of a quarter-scale model of the cask were conducted, and results were correlated with analytic predictions to verify structural and mechanical performance of the cask. Full-scale tests of the canisters were conducted to verify structural behavior of canister internals which provide criticality control. This paper describes the testing program for the NuPac 125-B Rail Cask, presents results therefrom, and correlates findings with Regulation 10 CFR 71 of the U.S. Nuclear Regulatory Commission

  20. Receptor-mediated sorting of soluble vacuolar proteins ends at the trans-Golgi network/early endosome.

    Science.gov (United States)

    Künzl, Fabian; Früholz, Simone; Fäßler, Florian; Li, Beibei; Pimpl, Peter

    2016-01-01

    The sorting of soluble proteins for degradation in the vacuole is of vital importance in plant cells, and relies on the activity of vacuolar sorting receptors (VSRs). In the plant endomembrane system, VSRs bind vacuole-targeted proteins and facilitate their transport to the vacuole. Where exactly these interactions take place has remained controversial, however. Here, we examine the potential for VSR-ligand interactions in all compartments of the vacuolar transport system in tobacco mesophyll protoplasts. To do this, we developed compartment-specific VSR sensors that assemble as a result of a nanobody-epitope interaction, and monitored the degree of ligand binding by analysing Förster resonance energy transfer using fluorescence lifetime imaging microscopy (FRET-FLIM). We show that VSRs bind ligands in the endoplasmic reticulum (ER) and in the Golgi, but not in the trans-Golgi network/early endosome (TGN/EE) or multivesicular late endosomes, suggesting that the post-TGN/EE trafficking of ligands towards the vacuole is VSR independent. We verify this by showing that non-VSR-ligands are also delivered to the vacuole from the TGN/EE after endocytic uptake. We conclude that VSRs are required for the transport of ligands from the ER and the Golgi to the TGN/EE, and suggest that the onward transport to the vacuole occurs by default. PMID:27249560

  1. CCC- and WASH-mediated endosomal sorting of LDLR is required for normal clearance of circulating LDL

    Science.gov (United States)

    Bartuzi, Paulina; Billadeau, Daniel D.; Favier, Robert; Rong, Shunxing; Dekker, Daphne; Fedoseienko, Alina; Fieten, Hille; Wijers, Melinde; Levels, Johannes H.; Huijkman, Nicolette; Kloosterhuis, Niels; van der Molen, Henk; Brufau, Gemma; Groen, Albert K.; Elliott, Alison M.; Kuivenhoven, Jan Albert; Plecko, Barbara; Grangl, Gernot; McGaughran, Julie; Horton, Jay D.; Burstein, Ezra; Hofker, Marten H.; van de Sluis, Bart

    2016-01-01

    The low-density lipoprotein receptor (LDLR) plays a pivotal role in clearing atherogenic circulating low-density lipoprotein (LDL) cholesterol. Here we show that the COMMD/CCDC22/CCDC93 (CCC) and the Wiskott–Aldrich syndrome protein and SCAR homologue (WASH) complexes are both crucial for endosomal sorting of LDLR and for its function. We find that patients with X-linked intellectual disability caused by mutations in CCDC22 are hypercholesterolaemic, and that COMMD1-deficient dogs and liver-specific Commd1 knockout mice have elevated plasma LDL cholesterol levels. Furthermore, Commd1 depletion results in mislocalization of LDLR, accompanied by decreased LDL uptake. Increased total plasma cholesterol levels are also seen in hepatic COMMD9-deficient mice. Inactivation of the CCC-associated WASH complex causes LDLR mislocalization, increased lysosomal degradation of LDLR and impaired LDL uptake. Furthermore, a mutation in the WASH component KIAA0196 (strumpellin) is associated with hypercholesterolaemia in humans. Altogether, this study provides valuable insights into the mechanisms regulating cholesterol homeostasis and LDLR trafficking. PMID:26965651

  2. Cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

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

  3. Cask fleet operations study

    International Nuclear Information System (INIS)

    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. Design and operational experience of dry cask storage systems

    International Nuclear Information System (INIS)

    This paper (Power Point presentation) describes cask storage design features and available dry cask storage technology, cask types used for dry storage, design characteristics of CASTOR casks, the German licensing basis for cask storage systems, shielding requirements, thermal layout, mechanical design, criticality safety and containment, licensing procedure, operational experience of dry cask storage in Germany and worldwide

  5. 78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-12-27

    ... RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS Cask... final rule amended the NRC's spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to...

  6. Derivation of sorting programs

    Science.gov (United States)

    Varghese, Joseph; Loganantharaj, Rasiah

    1990-01-01

    Program synthesis for critical applications has become a viable alternative to program verification. Nested resolution and its extension are used to synthesize a set of sorting programs from their first order logic specifications. A set of sorting programs, such as, naive sort, merge sort, and insertion sort, were successfully synthesized starting from the same set of specifications.

  7. Receptor-Mediated Uptake and Intracellular Sorting of Multivalent Lipid Nanoparticles Against the Epidermal Growth Factor Receptor (EGFR) and the Human EGFR 2 (HER2)

    Science.gov (United States)

    Tran, David Tu

    In the area of receptor-targeted lipid nanoparticles for drug delivery, efficiency has been mainly focused on cell-specificity, endocytosis, and subsequently effects on bioactivity such as cell growth inhibition. Aspects of targeted liposomal uptake and intracellular sorting are not well defined. This dissertation assessed a series of ligands as targeted functional groups against HER2 and EGFR for liposomal drug delivery. Receptor-mediated uptake, both mono-targeted and dual-targeted to multiple receptors of different ligand valence, and the intracellular sorting of lipid nanoparticles were investigated to improve the delivery of drugs to cancer cells. Lipid nanoparticles were functionalized through a new sequential micelle transfer---conjugation method, while the micelle transfer method was extended to growth factors. Through a combination of both techniques, anti-HER2 and anti-EGFR dual-targeted immunoliposomes with different combinations of ligand valence were developed for comparative studies. With the array of lipid nanoparticles, the uptake and cytotoxicity of lipid nanoparticles in relationship to ligand valence, both mono-targeting and dual-targeting, were evaluated on a small panel of breast cancer cell lines that express HER2 and EGFR of varying levels. Comparable uptake ratios of ligand to expressed receptor and apparent cooperativity were observed. For cell lines that express both receptors, additive dose-uptake effects were also observed with dual-targeted immunoliposomes, which translated to marginal improvements in cell growth inhibition with doxorubicin delivery. Colocalization analysis revealed that ligand-conjugated lipid nanoparticles settle to endosomal compartments similar to their attached ligands. Pathway transregulation and pathway saturation were also observed to affect trafficking. In the end, liposomes routed to the recycling endosomes were never observed to traffic beyond the endosomes nor to be exocytose like recycled ligands. Based on

  8. SNF shipping cask shielding analysis

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Pace, J.V. III

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan.

  9. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan

  10. GNS spent fuel cask experience

    International Nuclear Information System (INIS)

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

  11. GNS spent fuel cask experience

    Energy Technology Data Exchange (ETDEWEB)

    Weh, R. (Gesellschaft fuer Nuklear-Service mbH, Hannover (Germany))

    1993-05-01

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

  12. Initiatives in transport cask licensing

    International Nuclear Information System (INIS)

    The variations in research reactor fuel form, configuration, irradiation characteristics, and transport cask have required a substantial number of transport cask licensing actions associated with foreign research reactor spent fuel transportation. When compounded by limited time for shipment preparations, due to contract timing or delayed receipt of technical data, the number and timing of certifications has adversely impacted the ability of regulatory agencies to support intended shipping schedules. This issue was brought into focus at a april, 1998 meeting among DOE, the US Nuclear Regulatory Commission, and DOE's spent fuel transportation contractors. (author)

  13. Initiatives in transport cask license

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, John [NAC International, Aiken, SC (United States). Foreign Research Reactor Liaison]. E-mail: nacaiken@aol.com

    1998-07-01

    The variations in research reactor fuel form, configuration, irradiation characteristics, and transport cask have required a substantial number of transport cask licensing actions associated with foreign research reactor spent fuel transportation. When compounded by limited time for shipment preparations, due to contract timing or delayed receipt of technical data, the number and timing of certifications has adversely impacted the ability of regulatory agencies to support intended shipping schedules. This issue was brought into focus at a april, 1998 meeting among DOE, the US Nuclear Regulatory Commission, and DOE's spent fuel transportation contractors. (author)

  14. Low-cost/high-integrity waste casks

    International Nuclear Information System (INIS)

    The MOSAIK cast iron casks for storage and transportation of waste have the following advantages: much higher activity content with a lower total volume compared with concrete waste packages; good shielding in connection with automated filling or underwater loading techniques leads to dose exposure reduction of the operating personnel; high cask integrity guarantees a tight containment and makes an additional fixation of the waste in the cask cavity unnecessary; and the low serial production costs of cast iron casks and the resulting volume reduction using these casks lead to a cost advantage under German licensing conditions. 5 figures

  15. Source storage and transfer cask: Users Guide

    International Nuclear Information System (INIS)

    The storage and shield cask for the dual californium source is designed to shield and transport up to 3.7 mg (2 Ci) of 252Cf. 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

  16. k -Bitonic sort

    Institute of Scientific and Technical Information of China (English)

    高庆狮; 胡玥; 刘志勇

    1999-01-01

    A k-bitonic sort which generalizes the bitonic sort is proposed. The theorem of the bitonic sort, which merges two monotonic sequences into one order sequence, is extended into the theorem of k-bitonic sort. The k-bitonic sort merges K (=2k or 2k-1) monotonic sequences into one order sequence in steps, where k=[K/2] is an integer and k≥1. The k-bitonic sort is the Batcher’s bitonic sort when k=1.

  17. Cell sorting by deterministic cell rolling

    OpenAIRE

    Choi, Sungyoung; Karp, Jeffrey M.; Karnik, Rohit

    2011-01-01

    This communication presents the concept of “deterministic cell rolling”, which leverages transient cell-surface molecular interactions that mediate cell rolling to sort cells with high purity and efficiency in a single step.

  18. Rationalizing transport operations: The TN 24 transport storage cask approach

    International Nuclear Information System (INIS)

    The number of transports of spent fuel interim storage casks can be reduced by improved standardized cask design. Optimization of cask design is based on two main technological choices: shielding and spent fuel support basket design. The approaches to optimizing cask design to improve payload is described for the Transnucleaire TN24 family of dual purpose transport and storage casks. (author)

  19. Avi Sorting Network

    OpenAIRE

    Avinash Bansal; Kamal Gupta

    2012-01-01

    Sorting network is an abstract mathematical modelwhich can be used as a multiple-input, multiple-output switchingnetwork to sort the data in ascending or descending order [1].Sorting has been one of the most critical applications on parallelcomputing machines. Many classic textbooks on algorithms likeThomas H. Cormen, therefore consider this problem in greatdetail and list many sorting network for this purpose [2]. Thereare many sorting algorithms as the Bubble / Insertion sorter,Odd-Even sor...

  20. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

    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

  1. Development of high capacity transportable storage cask

    International Nuclear Information System (INIS)

    Mitsubishi Heavy Industries have developed high performance and reliable transportable storage casks, MSF series casks. The casks have employed newly developed materials that have been expressly developed to obtain long-term stability and quality. Furthermore, the casks have been employed newly designed structure to maximize payload of accommodating fuel assemblies in order to increase economic efficiency of storing spent fuels. The casks have been applied the following technologies. Basket assembly of the cask is made of newly developed boronated aluminum. The boronated aluminum is manufactured by power metallurgy process to provide uniformity of metallic structure and artificial aging which causes deterioration under high temperature condition is not applied to provide the boronated aluminum with high stability for long-term use. For the cask for BWR fuel, simplified basket whose design is that basket consists of some individual squire pipes without assembling is adopted in the cask. Neutron shielding material of the cask is made of newly resin of which raw materials have been modified to improve durability. Monolithic forging method which is how to shape steel into vessel form is developed to skip welding process between body shell and base plate and to improve reliability. Internal face of the body forging is machined to provide steps' in its cross section in order to fit the external shape of basket assembly and so heat dissipation performance is greatly improved. The new technologies have been done demonstration test in order to confirm that MSF series casks satisfy transport regulations. (author)

  2. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    International Nuclear Information System (INIS)

    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

  3. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Avi Sorting Network

    Directory of Open Access Journals (Sweden)

    Avinash Bansal

    2012-09-01

    Full Text Available Sorting network is an abstract mathematical modelwhich can be used as a multiple-input, multiple-output switchingnetwork to sort the data in ascending or descending order [1].Sorting has been one of the most critical applications on parallelcomputing machines. Many classic textbooks on algorithms likeThomas H. Cormen, therefore consider this problem in greatdetail and list many sorting network for this purpose [2]. Thereare many sorting algorithms as the Bubble / Insertion sorter,Odd-Even sorter, Sort the data in O(log2 n2 time complexity andsome other sorter have O(n2 as time complexity, where n is thenumber of elements. In this paper we propose a sorting networkcalled “Avi S orter” having time complexity O(n log2 n which isbased on just similar to bubble sort algorithm. This sortingnetwork provides the easy way to understand and manipulate theconcept of sorting network.

  5. Genetics Home Reference: CASK-related intellectual disability

    Science.gov (United States)

    ... Conditions CASK-related intellectual disability CASK-related intellectual disability Enable Javascript to view the expand/collapse boxes. ... Open All Close All Description CASK -related intellectual disability is a disorder of brain development that has ...

  6. Development of cask and transportation system

    International Nuclear Information System (INIS)

    Transportation of spent fuels to the AFR interim storage facility and disposal repository are necessary in Korea. Therefore, an emphasis has been concentrated to develop the design and fabrication technology of commercial casks. A conceptual design of the temperature and deformation measuring systems in the cask, which will be used for mock-up tests has been performed. Preliminary design data of the cask for 7 spent PWR fuels have been obtained in the course of study. (author)

  7. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

  8. Transportation capabilities of the existing cask fleet

    International Nuclear Information System (INIS)

    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 465 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the 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

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

    International Nuclear Information System (INIS)

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

  10. Used Fuel Cask Identification through Neutron Profile

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Laboratory

    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.

  11. Life cycle cost report of VHLW cask

    International Nuclear Information System (INIS)

    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

  12. What is a Sorting Function?

    DEFF Research Database (Denmark)

    Henglein, Fritz

    2009-01-01

    What is a sorting function—not a sorting function for a given ordering relation, but a sorting function with nothing given? Formulating four basic properties of sorting algorithms as defining requirements, we arrive at intrinsic notions of sorting and stable sorting: A function is a sorting funct...

  13. Roterende sorte huller

    DEFF Research Database (Denmark)

    Jerslev, Kristian

    Sorte huller har normalt været anset som statiske, mens alle andre legemer i Universet roterer. Dette stemmer imidlertid ikke overens med den nylige opdagelse af et roterende sort hul i midten af galaksen NGC 1365. Hvilken effekt har rotationen af et sort hul på legemer i dets nærhed, og hvordan...... kan astronomer i det hele taget måle, at sorte huller roterer?...

  14. Sorting Algorithms with Restrictions

    CERN Document Server

    Aslanyan, Hakob

    2011-01-01

    Sorting is one of the most used and well investigated algorithmic problem [1]. Traditional postulation supposes the sorting data archived, and the elementary operation as comparisons of two numbers. In a view of appearance of new processors and applied problems with data streams, sorting changed its face. This changes and generalizations are the subject of investigation in the research below.

  15. Identification of a novel protein-protein interaction motif mediating interaction of GPCR-associated sorting proteins with G protein-coupled receptors

    DEFF Research Database (Denmark)

    Bornert, Olivier; Møller, Thor Christian; Boeuf, Julien;

    2013-01-01

    degradation pathway. This protein belongs to the recently identified GPCR-associated sorting proteins (GASPs) family that comprises ten members for which structural and functional details are poorly documented. We present here a detailed structure-function relationship analysis of the molecular interaction...... GPCRs and highlight the presence within GASPs of a novel protein-protein interaction motif that might represent a new target to investigate the involvement of GASPs in the modulation of the activity of GPCRs.......GPCR desensitization and down-regulation are considered key molecular events underlying the development of tolerance in vivo. Among the many regulatory proteins that are involved in these complex processes, GASP-1 have been shown to participate to the sorting of several receptors toward the...

  16. Shielding Analysis of the 5320 Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, A. [Westinghouse Savannah River Company, AIKEN, SC (United States); Nathan, S. [Westinghouse Safety Management Solutions, Aiken, SC (United States)

    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.

  17. The C-terminal Extension of a Hybrid Immunoglobulin A/G Heavy Chain Is Responsible for Its Golgi-mediated Sorting to the Vacuole

    OpenAIRE

    Hadlington, Jane L.; Santoro, Aniello; Nuttall, James; Denecke, Jürgen; Ma, Julian K-C.; Vitale, Alessandro; Frigerio, Lorenzo

    2003-01-01

    We have assessed the ability of the plant secretory pathway to handle the expression of complex heterologous proteins by investigating the fate of a hybrid immunoglobulin A/G in tobacco cells. Although plant cells can express large amounts of the antibody, a relevant proportion is normally lost to vacuolar sorting and degradation. Here we show that the synthesis of high amounts of IgA/G does not impose stress on the plant secretory pathway. Plant cells can assemble ant...

  18. Design review report FFTF interim storage cask

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Design review report FFTF interim storage cask

    International Nuclear Information System (INIS)

    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

  20. 78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    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 System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 9...

  1. Sorting a distribution theory

    CERN Document Server

    Mahmoud, Hosam M

    2011-01-01

    A cutting-edge look at the emerging distributional theory of sorting Research on distributions associated with sorting algorithms has grown dramatically over the last few decades, spawning many exact and limiting distributions of complexity measures for many sorting algorithms. Yet much of this information has been scattered in disparate and highly specialized sources throughout the literature. In Sorting: A Distribution Theory, leading authority Hosam Mahmoud compiles, consolidates, and clarifies the large volume of available research, providing a much-needed, comprehensive treatment of the

  2. Seismic considerations for spent nuclear fuel storage in dry casks

    Institute of Scientific and Technical Information of China (English)

    John L Bignell; Jeffrey A Smith; Christopher A Jones; Susan Y Pickering

    2013-01-01

    To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems.Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters.The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g.A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping.In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask.The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over).The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask.Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.

  3. Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

    International Nuclear Information System (INIS)

    To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

  4. Fuel sorting evaluation

    International Nuclear Information System (INIS)

    An evaluation of functions and requirements associated with sorting fuel from the Hanford K Basins is presented to support design issue resolution decisions for achieving interim fuel storage. Potential requirements are recommended for implementation in design activities. The recommendations are provided as input to a management decision process where decisions are finalized and the sorting issue is closed

  5. Three Sorts of Naturalism

    DEFF Research Database (Denmark)

    Fink, Hans

    2006-01-01

    In "Two sorts of Naturalism" John McDowell is sketching his own sort of naturalism in ethics as an alternative to "bald naturalism". In this paper I distinguish materialist, idealist and absolute conceptions of nature and of naturalism in order to provide a framework for a clearer understanding o...

  6. Sorting to Extremes

    Science.gov (United States)

    Baum, Sandy; McPherson, Michael S.

    2011-01-01

    The world of higher education is a world of sorting, selecting, and ranking--on both sides of the market. Colleges select students to recruit and then to admit; students choose where to apply and which offer to accept. The sorting process that gets the most attention is in the higher reaches of the market, where it is not too much to say that…

  7. Layers in sorting practices: Sorting out patients with potential cancer

    DEFF Research Database (Denmark)

    Møller, Naja Holten; Bjørn, Pernille

    2011-01-01

    sorting mechanism, but is handled by informal sorting mechanisms. We identify two informal sorting mechanisms with large impact on the sorting practices, namely subtle categorizing and collective remembering. These informal sorting mechanisms have implications for the design of electronic booking systems...

  8. BR-100 spent fuel shipping cask development

    International Nuclear Information System (INIS)

    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 ampersand 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

  9. Transport and interim storage casks in Switzerland

    International Nuclear Information System (INIS)

    Full text: The Swiss utilities have chosen two different ways for the management of their spent fuel after initial on-site cooling: either reprocessing at La Hague plant (COGEMA) and Sellafield plant (COGEMA); or interim storage at the Central Interim Storage Facility called 'Zwischenlager Wuerenlingen AG' ( ZWILAG). Following international call for tenders, COGEMA LOGISTICS were awarded contracts for the supply of dual-purpose transport and storage casks for the interim storage of various spent fuel assemblies. All these casks belong to the family of the TN 24 dual purpose spent fuel storage casks in operation in the USA and in Belgium as well. They offer utilities a modular solution for the interim storage of spent fuel in robust metal casks which are fully suitable for off site transports. This flexible product can be readily adapted to suit individual user needs. The Leibstadt Nuclear Power Plant (KKL) has purchased six licensed dual-purpose TN97L spent fuel casks (97 BWR type fuel assemblies capacity). Three of them are already in operation at ZWILAG. COGEMA LOGISTICS has also delivered a dual-purpose TN52L spent fuel casks (52 BWR type fuel assemblies capacity) presently used for transport of spent fuel for reprocessing. The Goesgen Nuclear Power Plant (KKG) has purchased four licensed dual-purpose TN24G spent fuel casks (37 PWR type fuel assemblies capacity). They are all in operation at ZWILAG. The Muehleberg Nuclear Power Plant (BKW/KKM) has purchased 2 TN24BH spent fuel casks (69 BWR type fuel assemblies capacity). At the time of this abstract, cold trials are carried out involving the shuttle transport cask TN9/4 procured by COGEMA LOGISTICS as well. This paper will present the main features of these casks and the main steps of their development and implementation: 1) Main features of the casks: - The basic structure is a thick steel cylindrical forging with a welded on forged bottom and two forged steel lids. Containment and gamma shielding features of

  10. Design report for cask transportation equipment

    International Nuclear Information System (INIS)

    In Korea, the spent fuels stored in the spent fuel storage pools in the domestic nuclear power plants significantly affects the continuation of the power plant operation. To solve this problem, KAERI has developed KSC-4 spent fuel shipping cask, which can transport 4 PWR spent fuel assemblies. Besides the development of the cask, KAERI developed transportation equipment which needed to use of KSC-4 cask. These equipment consist of cask handling tools such as lifting yoke, lid handling tool and spent fuel handling tool, etc. and transportation equipment such as trailer. In this report the usages, structures and functions of these tools and equipment were described, and the safety evaluation was carried out for each equipment

  11. Cooling performance evaluation of the concrete cask

    International Nuclear Information System (INIS)

    The concrete cask storage system stores spent fuel by first sealing it within canisters and then containing such canisters inside a concrete cask. This report describes the results of a full-size model test performed to examine the heat dissipation characteristics of the concrete cask and to ascertain its ability to deal with elevated temperature. The specification to which a full-size concrete cask model was fabricated assumed an interim storage of 17x17UO2 fuel that was burned in PWR, estimating the heating value of spent fuel containing canister to be approximately 20 kW apiece. The test, which actually covered canister heating values ranging from 10 kW to 30 kW per unit to allow for temperature variations likely to be experienced in actual operation, verified that the concrete cask member did not exceed temperature limits. Test condition anticipated highest air temperature inside the spent fuel storage facility to be 30degC and, with reference to existing standard, set temperature limits of 65degC or less for the main body of concrete and 90degC or less for the local part as criteria. Preliminary 3-D thermo hydrodynamic analysis done prior to the test indicated that the temperature of the local part of the concrete cask member would be below 90degC. It also confirmed that steel material used as the structural member of the canisters or concrete cask would remain around 200degC even in an area where it was highest, validating that the integrity of such material would pose no problem from the analytical point of view. Heat dissipation performance test conducted in steady state verified that the concrete cask was able to have a sufficient cooling capacity against per-canister heating values in the 10 kW to 30 kW range which had been chosen in anticipation of temperature variation thought to be encountered in actual service. Also, to examine the consequence of the concrete cask having lost its cooling ability, another heat dissipation test was carried out under

  12. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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

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

    International Nuclear Information System (INIS)

    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 fuer 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

  14. SNS Inner Plug Shipping Cask Analysis

    International Nuclear Information System (INIS)

    Calculations were performed to evaluate the dose rates outside the shipping cask containing the Spallation Neutron Source (SNS) inner plug assembly. The analysis consisted of simulating the proton beam interaction with the SNS target, activation calculations with the determined neutron flux levels and assumed SNS operation schedule, and calculation of the decay gamma-rays propagation through the inner plug and shipping cask. Several materials were considered for the inner plug. The results provide guidance for the finalization of the plug design

  15. A cask maintenance facility feasibility study

    International Nuclear Information System (INIS)

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

  16. Cask Processing Enclosure Specification/Operation - 12231

    International Nuclear Information System (INIS)

    Following an evaluation of throughput rates in the Hot Cell at the Transuranic Waste Processing Center and considering the variability in the waste with respect to actual dose rates a new approach to processing transuranic waste was necessary. Compounding the issue was the remote equipment poor reliability and high down-time. After considering all the factors, the evaluation resulted in the design and construction of a new waste processing area for handling the concrete casks that predominately contain contact handled transuranic (TRU) waste. The area is called the Cask Processing Enclosure and essentially the Cask Processing Enclosure mimics the projects current process techniques used for processing Contact Handled -TRU waste in the existing Box Breakdown Area and Glovebox. The Cask Processing Enclosure approach was developed based on a review of the RH processing throughput rates in the Hot Cell. As the process was reviewed consideration was given to the variability in the waste with respect to actual dose rates and the lack of equipment reliability and high wear in the Hot Cell. Based on that review, a new contact handled processing area for handling the concrete casks is being constructed and startup is expected shortly following WM2012. The Cask Processing Enclosure essentially mimics the projects current process techniques used for processing Contact Handled waste in the existing Box Breakdown Area and Glovebox and the design takes into consideration six years of operational experience. (authors)

  17. Wage Sorting Trends

    DEFF Research Database (Denmark)

    Bagger, Jesper; Vejlin, Rune Majlund; Sørensen, Kenneth Lykke

    Using a population-wide Danish Matched Employer-Employee panel from 1980-2006, we document a strong trend towards more positive assortative wage sorting. The correlation between worker and firm fixed effects estimated from a log wage regression increases from -0.07 in 1981 to .14 in 2001. The...... nonstationary wage sorting pattern is not due to compositional changes in the labor market, primarily occurs among high wage workers, and comprises 41 percent of the increase in the standard deviation of log real wages between 1980 and 2006. We show that the wage sorting trend is associated with worker...

  18. 78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc.... Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel...

  19. Sorting Out Seasonal Allergies

    Science.gov (United States)

    ... Close ‹ Back to Health Library Sorting Out Seasonal Allergies Sneezing, runny nose, nasal congestion. Symptoms of the ... How do I know if I have seasonal allergies? According to Dr. Georgeson, the best way to ...

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

    International Nuclear Information System (INIS)

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

  1. Advanced handling technology project and implications for cask design

    International Nuclear Information System (INIS)

    This paper describes the results of the ongoing Advanced Handling Technologies Project (AHTP) at Sandia. AHTP was initiated in 1986 to explore the use of advanced robotic systems to perform cask handling operations at radioactive waste handling facilities and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof of concept systems developed in AHTP are intended to extrapolate from currently available commercial systems to those that would be available by the time than an actual repository would be open for operation. These systems provide test facilities for the investigation of the robotic handling of alternate cask design features. The following sections describe (1) the approach used in AHTP to select operations for proof of concept robotic systems and to identify the cask design implications, (2) the separate proof of concept robotic systems developed in AHTP, and (3) preliminary insights into the impact of cask system design features on the feasibility of robotic performance of cask handling operations. The main conclusions from AHTP to date regarding design for remote handling are: (1) incorporation of cask system design features which facilitate robotic cask handling can be achieved with minimal impact on cask functional features, (2) proper cask design allows robotic cask handling operations from manipulation of cask tie-down mechanisms to radiation surveys to be performed quickly and reliably without direct human intervention, and (3) design for remote handling also facilitates manual handling operations

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

    International Nuclear Information System (INIS)

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

  3. Full scale torch tests on spent fuel cask shipping system

    International Nuclear Information System (INIS)

    Full scale experimental measurements, including the instrumentation designed to obtain the data, are presented on the thermal effects of torch fires on a large, spent nuclear fuel shipping cask. The measured temperature data in the various materials of the multilayered cask are unique, since no torch tests have been previously performed on a cask: These data were obtained during a series of four torch tests which simulate a situation in which the relief valve of a liquefied gas tank railcar has been opened and and the contents are vented and ignited so that the resultant torch impinges on the cask. The modified cask instrumentation geometry and materials are discussed. Temperature data throughout the cask are compared for two cask on the corrugated outer jacket surface, within the neutron shield, on the carbon steel shell, on the inner stainless steel shell and near the cask head closure seals are presented for the four torch tests

  4. Multiple pathways for vacuolar sorting of yeast proteinase A

    DEFF Research Database (Denmark)

    Westphal, V; Marcusson, E G; Winther, Jakob R.; Emr, S D; van den Hazel, H B

    1996-01-01

    The sorting of the yeast proteases proteinase A and carboxypeptidase Y to the vacuole is a saturable, receptor-mediated process. Information sufficient for vacuolar sorting of the normally secreted protein invertase has in fusion constructs previously been found to reside in the propeptide of...

  5. Analysis of DCI cask drop test onto reinforced concrete pad

    International Nuclear Information System (INIS)

    In a cask-storage facility, a cask may be subjected to an impact load as a result of a free drop onto the floor because of cask mishandling. We performed drop tests of casks onto a reinforced concrete (RC) slab representing the floor of a facility as well as simulation analysis [Kato et al]. This paper describes the details of the FEM analysis and calculated results and compares them with the drop test results. (J.P.N.)

  6. Spent Fuel Dry Storage Cask Thermal Test

    International Nuclear Information System (INIS)

    Most nuclear power plants maintain their spent fuel discharged at a reactor in wet storage pools. However, after several years of use, many pools are filled to capacity. Therefore, finding a sufficient capacity for storage is essential because of the continued delays in obtaining a safe, interim storage facility if nuclear power plants are to be allowed to continue to operate. Dry storage cask will be one solution for solving an interim storage problem. The dry storage cask consists of two separate components: an over-pack, and a canister. The structure strength part of the over-pack is made of carbon steel, and the inner cavity of the structure strength part is filled with concrete, which accomplishes the role as a radiation shield. The outer diameter of the dry storage cask is 3,550 mm and the its overall height is 5,885 mm. It weighs approximately 135 tons. The dry storage cask accommodates 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat from the 24 PWR spent fuel assemblies is 25.2 kW This paper discusses the experimental approach used to evaluate the heat transfer characteristics of the dry storage cask

  7. Ready, steady, SORT!

    CERN Multimedia

    Laëtitia Pedroso

    2010-01-01

    The selective or ecological sorting of waste is already second nature to many of us and concerns us all. As the GS Department's new awareness-raising campaign reminds us, everything we do to sort waste contributes to preserving the environment.    Placemats printed on recycled paper using vegetable-based ink will soon be distributed in Restaurant No.1.   Environmental protection is never far from the headlines, and CERN has a responsibility to ensure that the 3000 tonnes and more of waste it produces every year are correctly and selectively sorted. Materials can be given a second life through recycling and re-use, thereby avoiding pollution from landfill sites and incineration plants and saving on processing costs. The GS Department is launching a new poster campaign designed to raise awareness of the importance of waste sorting and recycling. "After conducting a survey to find out whether members of the personnel were prepared to make an effort to sort a...

  8. Microfluidic sorting of microtissues.

    Science.gov (United States)

    Buschke, D G; Resto, P; Schumacher, N; Cox, B; Tallavajhula, A; Vivekanandan, A; Eliceiri, K W; Williams, J C; Ogle, B M

    2012-03-01

    Increasingly, invitro culture of adherent cell types utilizes three-dimensional (3D) scaffolds or aggregate culture strategies to mimic tissue-like, microenvironmental conditions. In parallel, new flow cytometry-based technologies are emerging to accurately analyze the composition and function of these microtissues (i.e., large particles) in a non-invasive and high-throughput way. Lacking, however, is an accessible platform that can be used to effectively sort or purify large particles based on analysis parameters. Here we describe a microfluidic-based, electromechanical approach to sort large particles. Specifically, sheath-less asymmetric curving channels were employed to separate and hydrodynamically focus particles to be analyzed and subsequently sorted. This design was developed and characterized based on wall shear stress, tortuosity of the flow path, vorticity of the fluid in the channel, sorting efficiency and enrichment ratio. The large particle sorting device was capable of purifying fluorescently labelled embryoid bodies (EBs) from unlabelled EBs with an efficiency of 87.3% ± 13.5%, and enrichment ratio of 12.2 ± 8.4 (n = 8), while preserving cell viability, differentiation potential, and long-term function. PMID:22505992

  9. Sorting Plastic Waste in Hydrocyclone

    Directory of Open Access Journals (Sweden)

    Ernestas Šutinys

    2011-02-01

    Full Text Available The article presents material about sorting plastic waste in hydrocyclone. The tests on sorting plastic waste were carried out. Also, the findings received from the performed experiment on the technology of sorting plastic waste are interpreted applying an experimental model of the equipment used for sorting plastics of different density.Article in Lithuanian

  10. NUHOMS registered - MP197 transport cask

    International Nuclear Information System (INIS)

    The NUHOMS registered -MP197 cask is an optimized transport design which can be loaded in the spent fuel pool (wet loading) or loaded the canister from the NUHOMS concrete modules at the ISFSI site. With impact limiters attached, the package can be transported within the states or world-wide. The NUHOMS registered -MP197 packaging can be used to transport either BWR or PWR canisters. The NUHOMS registered -MP197 cask is designed to the ASME B and PV Code and meets the requirements of Section III, Division 3 for Transport Packaging. The cask with impact limiters has undergone drop testing to verify the calculated g loadings during the 9m drops. The test showed good correlation with analytical results and demonstrate that the impact limiters stay in place and protect the package and fuel during the hypothetical accidents

  11. 78 FR 73379 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... of approved spent fuel storage casks in 10 CFR 72.214 (65 FR 25241; May 1, 2000). The environmental... 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9...

  12. Economic evaluation of nuclear waste transportation casks

    International Nuclear Information System (INIS)

    A method is described which allows the systematic economic evaluation of transportation cask designs which meet the requirements of the Test and Evaluation Facility (TEF) program. The heart of the method described is the Waste Management Transportation Model. This model uses a set of computer-based algorithms to assemble specific case information input, combine this input with the data base of transportation information maintained within the model, and calculate the cask types and quantities necessary, the cask utilization factors, and the total costs for each transport line specified. The model is capable of handling a large variety of transportation problems given the specific input related to each type. Three combinations of waste packaging facilities were examined. The first assumes all consolidation and packaging occurs at an existing hot cell. The second assumes all consolidation and packaging is done at the TEF site. The third combination assumes that spent fuels are consolidated at an existing hot cell while waste packaging occurs at the TEF site. Some of the general findings are: (1) defense high-level waste (DHLW) is generally lower in cost than SF as the prime waste form because of the fewer number of shipments required prior to the waste consolidation activity; (2) when DHLW is the prime waste form, it is beneficial to locate the packaging facility (PF) close to the TEF site because the packaged waste form is heavier, more costly to transport; (3) when SF is the prime waste form, it is beneficial to locate the PF close to the waste source to reduce the length of the transport links containing unconsolidated spent fuel assemblies; and (4) truck casks, and legal weight truck casks in particular, are generally superior to the rail casks on an economic basis

  13. Hexagonal absorption cask for nuclear power

    International Nuclear Information System (INIS)

    A hexagonal absorption cask for compact spent fuel storage is designed. The cask is made of austenitic stainless steel with a high boron content. One of the two sides of each of the six wall plates is longitudinally chamfered and attached to the inner face of the next wall plate in the hexagonal arrangement. The whole is welded together. This design secures that the absorption of the neutron flux in the radial direction will not be deteriorated if the boron content of the weld metal is reduced. (Z.S.). 2 figs

  14. Safety evaluation for packaging (onsite) SERF cask

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

    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

  16. Status of the Beneficial Uses Shipping System cask (BUSS)

    International Nuclear Information System (INIS)

    The Beneficial Uses Shipping System cask is a Type B packaging developed by Sandia National Laboratories for the U.S. Department of Energy. The cask is designed to transport special form radioactive source capsules (cesium chloride and strontium fluoride) produced by the Department of Energy's Hanford Waste Encapsulation and Storage Facility. This paper describes the cask system and the analyses performed to predict the response of the cask in impact, puncture, and fire accident conditions as specified in the regulations. The cask prototype has been fabricated and Certificates of Compliance have been obtained

  17. A numerical study of transportation casks subjected to puncture loads

    International Nuclear Information System (INIS)

    A nonlinear dynamic finite element analysis has been performed to study the structural response of casks subjected to puncture load. Particular attention is placed on the Multipurpose Canister (MPC) and General Atomic (GA) casks that are currently under development. The structural response of the casks subjected to both regulatory hypothetical accidents and accidents beyond regulatory requirements were evaluated. A performance map was presented for casks subjected to regulatory formula puncture tests, and the structural contribution of the various layers backing the steel cask shell has been studied

  18. Sorting and sustaining cooperation

    DEFF Research Database (Denmark)

    Vikander, Nick

    2013-01-01

    This paper looks at cooperation in teams where some people are selfish and others are conditional cooperators, and where lay-offs will occur at a fixed future date. I show that the best way to sustain cooperation prior to the lay-offs is often in a sorting equilibrium, where conditional cooperators...

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

    International Nuclear Information System (INIS)

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

  20. Seismic stability of unanchored spent nuclear fuel storage casks

    International Nuclear Information System (INIS)

    Dynamic soil-structure interaction analyses were performed to examine the effects of a potential earthquake on the stability of unanchored cylindrical spent nuclear fuel casks for an above-ground storage installation. The casks would be placed on a cluster of reinforced concrete pads founded on a deep sequence of clays and silts underlain by sandstones. The analyses focused on evaluating the geometric stability of the casks during an earthquake with respect to a design concept that a cask should not tip over, slide off the storage pad, or collide with another cask. The analyses were performed using LS-DYNA with a three-dimensional explicit finite element model representing the site soil and a fully loaded storage pad. Three statistically independent acceleration time histories were applied simultaneously at the base of the model to generate a free-field ground motion time history representing the design-basis earthquake. Sensitivity studies were performed to examine the effects of the interface conditions between the storage pad and the surrounding soil, and between the base of the storage casks and the top surface of the pad. The results indicate that ground motion from the design-basis earthquake would not cause any cask to tip over, slide off the pad, or collide with another cask. The contact conditions at the cask-to-pad and pad-to-soil interfaces have a strong effect on potential cask motions during an earthquake. If the cask-base friction coefficient is small, the casks may slide, but would not experience any significant rocking. If the cask-base friction is large enough to permit a significant transfer of earthquake lateral motions across the cask-to-pad interface, a design with bonded pad-to-soil interfaces would produce larger cask motions than a design with frictional pad-to-soil interfaces. Furthermore, a cask strage design in which the cask motions are essentially isolated from the motions of the pad-soil system, which can be accomplished if the cask

  1. 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)

    International Nuclear Information System (INIS)

    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

  2. Simplified computer codes for cask impact analysis

    International Nuclear Information System (INIS)

    In regard to the evaluation of the acceleration and deformation of casks, the simplified computer codes make analyses economical and decrease input and calculation time. The results obtained by the simplified computer codes have enough adequacy for their practical use. (J.P.N.)

  3. Surface storage cask test summarization report

    International Nuclear Information System (INIS)

    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

  4. Sorting quantum systems efficiently

    Science.gov (United States)

    Ionicioiu, Radu

    2016-01-01

    Measuring the state of a quantum system is a fundamental process in quantum mechanics and plays an essential role in quantum information and quantum technologies. One method to measure a quantum observable is to sort the system in different spatial modes according to the measured value, followed by single-particle detectors on each mode. Examples of quantum sorters are polarizing beam-splitters (PBS) – which direct photons according to their polarization – and Stern-Gerlach devices. Here we propose a general scheme to sort a quantum system according to the value of any d-dimensional degree of freedom, such as spin, orbital angular momentum (OAM), wavelength etc. Our scheme is universal, works at the single-particle level and has a theoretical efficiency of 100%. As an application we design an efficient OAM sorter consisting of a single multi-path interferometer which is suitable for a photonic chip implementation. PMID:27142705

  5. Gold ore sorting

    International Nuclear Information System (INIS)

    Apparatus for sorting lumps of gold-bearing ore according to their gold content is described. It includes means for irradiating the lumps of ore with neutrons, e.g. a neutron tube adapted to produce at least 1010 neutrons per second with an energy of less than 4.5 MeV. The resulting intensity of 297 keV gamma rays arising from the nuclear reaction 197Au(n,n'#betta#) 197Au is measured. The measured gamma ray intensity from a given lump of ore is used to sort that lump of ore from other lumps. The apparatus includes various cylinders and a vibrator for presenting the lumps of ore to the neutrons in a geometrical configuration such as to enable the lumps to be irradiated uniformly. (author)

  6. Lipoprotein sorting in bacteria.

    Science.gov (United States)

    Okuda, Suguru; Tokuda, Hajime

    2011-01-01

    Bacterial lipoproteins are synthesized as precursors in the cytoplasm and processed into mature forms on the cytoplasmic membrane. A lipid moiety attached to the N terminus anchors these proteins to the membrane surface. Many bacteria are predicted to express more than 100 lipoproteins, which play diverse functions on the cell surface. The Lol system, composed of five proteins, catalyzes the localization of Escherichia coli lipoproteins to the outer membrane. Some lipoproteins play vital roles in the sorting of other lipoproteins, lipopolysaccharides, and β-barrel proteins to the outer membrane. On the basis of results from biochemical, genetic, and structural studies, we discuss the biogenesis of lipoproteins in bacteria, their importance in cellular functions, and the molecular mechanisms underlying efficient sorting of hydrophobic lipoproteins to the outer membrane through the hydrophilic periplasm. PMID:21663440

  7. Sorting quantum systems efficiently.

    Science.gov (United States)

    Ionicioiu, Radu

    2016-01-01

    Measuring the state of a quantum system is a fundamental process in quantum mechanics and plays an essential role in quantum information and quantum technologies. One method to measure a quantum observable is to sort the system in different spatial modes according to the measured value, followed by single-particle detectors on each mode. Examples of quantum sorters are polarizing beam-splitters (PBS) - which direct photons according to their polarization - and Stern-Gerlach devices. Here we propose a general scheme to sort a quantum system according to the value of any d-dimensional degree of freedom, such as spin, orbital angular momentum (OAM), wavelength etc. Our scheme is universal, works at the single-particle level and has a theoretical efficiency of 100%. As an application we design an efficient OAM sorter consisting of a single multi-path interferometer which is suitable for a photonic chip implementation. PMID:27142705

  8. Heideggers sorte arv

    DEFF Research Database (Denmark)

    Olesen, Søren Gosvig

    2015-01-01

    Martin Heidegger var antisemit, men er hans tænkning og intellektuelle arv det også? Søren Gosvig Olesen opsøger den store tyske tænkers arvinger og bindene fra 1938-48 i Heideggers efterladte ’Sorte hæfter’, hvor den lille mands meninger blander sig med en stor tænkers tanker......Martin Heidegger var antisemit, men er hans tænkning og intellektuelle arv det også? Søren Gosvig Olesen opsøger den store tyske tænkers arvinger og bindene fra 1938-48 i Heideggers efterladte ’Sorte hæfter’, hvor den lille mands meninger blander sig med en stor tænkers tanker...

  9. Event shape sorting

    Science.gov (United States)

    Kopečná, Renata; Tomášik, Boris

    2016-04-01

    We propose a novel method for sorting events of multiparticle production according to the azimuthal anisotropy of their momentum distribution. Although the method is quite general, we advocate its use in analysis of ultra-relativistic heavy-ion collisions where a large number of hadrons is produced. The advantage of our method is that it can automatically sort out samples of events with histograms that indicate similar distributions of hadrons. It takes into account the whole measured histograms with all orders of anisotropy instead of a specific observable ( e.g., v_2 , v_3 , q_2 . It can be used for more exclusive experimental studies of flow anisotropies which are then more easily compared to theoretical calculations. It may also be useful in the construction of mixed-events background for correlation studies as it allows to select events with similar momentum distribution.

  10. Chip-based droplet sorting

    Energy Technology Data Exchange (ETDEWEB)

    Beer, Neil Reginald; Lee, Abraham; Hatch, Andrew

    2014-07-01

    A non-contact system for sorting monodisperse water-in-oil emulsion droplets in a microfluidic device based on the droplet's contents and their interaction with an applied electromagnetic field or by identification and sorting.

  11. Chip-based droplet sorting

    Science.gov (United States)

    Beer, Neil Reginald; Lee, Abraham; Hatch, Andrew

    2014-07-01

    A non-contact system for sorting monodisperse water-in-oil emulsion droplets in a microfluidic device based on the droplet's contents and their interaction with an applied electromagnetic field or by identification and sorting.

  12. Teaching Sorting in ICT

    OpenAIRE

    Szlávi, Péter; Törley, Gábor

    2009-01-01

    This article is aimed at considering how an algorithmic problem - more precisely a sorting problem - can be used in an informatics class in primary and secondary education to make students mobilize the largest possible amount of their intellectual skills in the problem solving process. We will be outlining a method which essentially forces students to utilize their mathematical knowledge besides algorithmization in order to provide an efficient solution. What is more, they are expected to ...

  13. Safety analysis of casks under extreme impact conditions

    International Nuclear Information System (INIS)

    The determination of the inherent safety of casks also under extreme impact conditions has been of increasing interest since the terrorist attacks from 11th September 2001. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. One of the most critical scenarios for a cask is the centric impact of a dynamic load onto the lid seal system. This can be caused e.g. by direct aircraft crash or its engine as well as by an impact due to the collapse of a building e.g. a nuclear facility storage hall. In this context BAM is developing methods to calculate the deformation of cask components and - with respect to leak tightness - relative displacements between the metallic seals and their counterparts. This paper presents reflections on modelling of cask structures for Finite Element analyses and discusses calculated results of stresses and deformations. Another important aspect is the behaviour of a cask under a lateral impact by aircraft and fragments of a building. Examples of the kinetic reaction (cask acceleration due to the fragments, subsequent contact with neighbouring structures like ground, buildings or casks) are shown and discussed in correlation to cask stresses which are to be expected

  14. Sorting out Ideas about Function

    Science.gov (United States)

    Hillen, Amy F.; Malik, LuAnn

    2013-01-01

    Card sorting has the potential to provide opportunities for exploration of a variety of topics and levels. In a card-sorting task, each participant is presented with a set of cards--each of which depicts a relationship--and is asked to sort the cards into categories that make sense to him or her. The concept of function is critical to…

  15. Pair Wise Sorting: A New Way of Sorting

    Directory of Open Access Journals (Sweden)

    Md. Jahangir Alam

    2010-12-01

    Full Text Available This paper presents a technique for sorting numerical data in an efficient way. The numbers of comparisons i.e. the running time of this technique is dependent on distribution or diversity of the value of data items as like as other efficient algorithms. When the total number of data is even, this method groups that data into a collection of pairs and therefore establishes the sorting constraints on each of the pairs. The control is traversed through the list of elements by changing the position of each pair which is the major principle of this technique. On the other hand, when the total number of elements is odd, this method sorts all elements except the last one in the same was as mentioned earlier and the last element is sorted using the general Insertion Sort. This algorithm is therefore a hybrid sorting method that sorts elementary numeric data in a faster and efficient manner.

  16. Evaluation of improvement potential for spent fuel cask handling

    International Nuclear Information System (INIS)

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

  17. Cask containing method for spent fuel assembly and subcriticality measuring device for a cask containing system

    International Nuclear Information System (INIS)

    An area for a spent fuel storage pool is sectioned into an ordinary rack area for disposing spent fuel assemblies taken out from a reactor core and a preliminary storage rack area having the same constitution as a cask for containing spent fuel assemblies. Preceding to cask-containment, the spent fuel assemblies are temporarily transferred once in the preliminary storing rack area from the ordinary rack area to ensure subcriticality and then contained in casks. In addition, those fuels having a higher burn-up degree are disposed coaxially to the central portion and those having not higher burn-up degree are disposed at the outer circumferential portion. The spent fuel assemblies can surely be contained in the casks, or the process of containing the spent fuel assemblies to the casks or the subcriticality after the containment can be evaluated thereby capable of further ensuring the subcriticality. The spent fuel assemblies can be transferred or stored safely and reliably at a good efficiency. (N.H.)

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

    International Nuclear Information System (INIS)

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

  19. Fluorescence activated cell sorting.

    Science.gov (United States)

    Bonner, W. A.; Hulett, H. R.; Sweet, R. G.; Herzenberg, L. A.

    1972-01-01

    An instrument has been developed for sorting biological cells. The cells are rendered differentially fluorescent and incorporated into a small liquid stream illuminated by a laser beam. The cells pass sequentially through the beam, and fluorescent light from the cells gives rise to electrical signals. The stream is broken into a series of uniform size drops downstream of the laser. The cell signals are used to give appropriate electrostatic charges to drops containing the cells. The drops then pass between two charged plates and are deflected to appropriate containers. The system has proved capable of providing fractions containing large numbers of viable cells highly enriched in a particular functional type.

  20. Inspection of Used Fuel Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Drop test of transportable storage cask

    International Nuclear Information System (INIS)

    It is being planned to transport the transportable storage casks again after their storage period of several decades, so metal gaskets are used as seal material in their lids in place of rubber o-rings which deteriorate during the storage period. Since the slightest dislocation of the lids causes seal performance deterioration in the metal gaskets, it is necessary to establish a simulation technology which accurately estimates the dislocation in order to design a rigid lid structure to protect against the impact loads under 9 m drop condition. A 1:3 scale model of the transportable storage cask developed by Hitz for BWR spent fuel rods were manufactured and 9 m drop tests were performed. Measured dislocations of the lids were confirmed within the allowable limit and they were found to be accurately simulated. (author)

  2. Transportation cask decontamination and maintenance at the potential Yucca Mountain repository

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  4. Interfacing the existing cask fleet with the MRS

    International Nuclear Information System (INIS)

    This paper reports that the Department of Energy (DOE) is considering the possibility of using the existing fleet of casks to achieve spent fuel receipt at the Monitored Retrievable Storage (MRS) facility. The existing cask fleet includes the NLI-1/2, the NAC-LWT, the TN-8 (and TN-8L), the TN-9, and the IF-300 casks. Other casks may be available, but their status is not certain. Use of the existing cask fleet at the MRS places additional design requirements on the system, and specifically affects the cask-to-MRS interface. The decision to use the existing cask fleet also places additional demands on training needs and operator certification, and the configuration management system. Some existing cask designs may not be able to mate with a bottom opening hot cell MRS. Use of the existing cask fleet also greatly increases the number of shipments that must be received, to the point that a facility larger than originally envisioned may be required

  5. Facility for the decontamination of fuel element transport casks

    International Nuclear Information System (INIS)

    Before transporting them to the reprocessing plant the transport casks containing fuel assemblies are cleaned on the outside. This is achieved by means of a tank which can be closed and into which the transport cask is inserted. Within the tank the cask is standing on a roller drive mechanism with the aid of which it can be revolved. In the bottom, shell, and cover region there are adjustable nozzles by means of which the transport cask can be sprayed with washing water all-round. After cleaning, drying is performed at subatmospheric pressure by means of stationary hot-air nozzles also arranged in the tank. (DG)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 TNTM81 casks currently in use in Switzerland and the TNTM85 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 TNTM81 and TNTM85 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 TNTM81 and the TNTM85 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 TNTM28 cask and TS 28 V cask, all along the logistics chain of loading, unloading, transport and maintenance. In addition, years of feedback and experience in design and operations - together with ever improved materials - have allowed finding further optimization of this type of cask design. In order to increase the loading capacity in terms of radioactive source terms and heat load by 40%, the cask design relies on innovative solutions and benchmarks from the current shipping campaigns. Currently, TNTM81 and TNTM85 are the only licensed casks that can transport and store 28 canisters with a total decay heat of 56 kW. It contributes to optimise the number of required transports to bring back high level waste residues to their producers. Three units have already been loaded and transported to

  8. Rail tiedown tests with heavy casks for radioactive shipments

    International Nuclear Information System (INIS)

    A rail tiedown test program was conducted at the Savannah River Plant in July and August 1978. For each test, a 40- or 70-ton cask was secured on a railcar. The railcar was pushed to speeds up to 11 mph and allowed to couple to parked railcars simulating ordinary railyard operations. The test car carrying the cask was heavily instrumented to measure the accelerations and forces generated at strategically selected places. Eighteen test runs were made with different combinations of railcars, couplers, casks, speeds, and tiedown configurations. The major objectives of the test program were to (1) provide test data as a basis to develop a tiedown standard for rail cask shipments of radioactive materials and (2) collect dynamic data to support analytical models of the railcar cask tiedown system. The optimum tiedown configuration demonstrated for heavy casks was a combination of welded, fixed stops to secure the cask longitudinally and flexible cables to restrain vertical and lateral cask movement. Cables alone were inadequate to secure a heavy cask to a standard railcar, and bolting was found disadvantageous in several respects. The use of cushioning coupler mechanisms dramatically reduced the tiedown requirements for the rail coupling operation. The test program and general conclusions are discussed

  9. MCO loading and cask loadout technical manual

    International Nuclear Information System (INIS)

    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. MCO loading and cask loadout technical manual

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Concrete Spent Fuel Cask Criticality Calculation

    International Nuclear Information System (INIS)

    A preliminary analysis of the concrete cask for the intermediate dry storage of the spent fuel of NPP Krsko should include an estimation of the effective multiplication factor. Assuming 16x16 fuel elements, 4.3% initial enrichment, 45 GWd/tU burnup and 10 years cooling time, a concrete spent fuel capacity of 10 spent fuel assemblies is proposed. Fuel assemblies are placed inside inner cavity in a 'basket' - a boron (1%) doped steel structure. Heavy concrete (25% Fe), 45 cm thick, is enclosed in a carbon steel shell. There is also a stainless steel (SS304) lining of the storage cavity. Isotope inventory of the spent fuel after a 10 years cooling time is calculated using ORIGEN-S functional module of the SCALE-4.2 code package. The effective multiplication factor keff of dry (helium filled) and wet (water filled) cask for fresh and used fuel is calculated using CSAS4 Monte Carlo method based control module of the same SCALE-4.2 code package. The obtained results of keff of the dry cask for fresh and spent fuel are well below the required 0.95 value, but those for the water filled cask are above this value. Therefore, several additional calculations of the keff varying the thickness of a boral basket structure which had replaced the stainless steel one were done. It turned out that at least a 1.5 cm thick boral wall was needed to meet the required 0.95 value for keff. (author)

  12. Selective sorting of waste

    CERN Multimedia

    2007-01-01

    Not much effort needed, just willpower In order to keep the cost of disposing of waste materials as low as possible, CERN provides two types of recipient at the entrance to each building: a green plastic one for paper/cardboard and a metal one for general refuse. For some time now we have noticed, to our great regret, a growing negligence as far as selective sorting is concerned, with, for example, the green recipients being filled with a mixture of cardboard boxes full of polystyrene or protective wrappers, plastic bottles, empty yogurts pots, etc. …We have been able to ascertain, after careful checking, that this haphazard mixing of waste cannot be attributed to the cleaning staff but rather to members of the personnel who unscrupulously throw away their rubbish in a completely random manner. Non-sorted waste entails heavy costs for CERN. For information, once a non-compliant item is found in a green recipient, the entire contents are sent off for incineration rather than recycling… We are all concerned...

  13. Chromosome analysis and sorting

    Czech Academy of Sciences Publication Activity Database

    Doležel, Jaroslav; Kubaláková, Marie; Suchánková, Pavla; Kovářová, Pavlína; Bartoš, Jan; Šimková, Hana

    Weinheim : Wiley-VCH, 2007 - (Doležel, J.; Greilhuber, J.; Suda, J.), s. 373-403 ISBN 978-3-527-31487-4 R&D Projects: GA ČR GA521/04/0607; GA ČR GP521/05/P257; GA ČR GD521/05/H013; GA MŠk(CZ) LC06004 Grant ostatní: Mendelova zemědělská a lesnická univerzita v Brně / Agronomická fakulta(CZ) ME 844 Institutional research plan: CEZ:AV0Z5038910 Source of funding: V - iné verejné zdroje ; V - iné verejné zdroje ; V - iné verejné zdroje ; V - iné verejné zdroje ; V - iné verejné zdroje Keywords : Plant flow cytometry * chromosome sorting * flow cytogenetics Subject RIV: EB - Genetics ; Molecular Biology http://books. google .com/books?id=3cwakORieqUC&pg=PA373&lpg=PA373&dq=Chromosome+analysis+and+sorting&source=web&ots=8IyvJlBQyq&sig=_NlXyQQgBCwpj1pTC9YITvvVZqU

  14. Dry storage cask - DIORIT - Swiss experience

    International Nuclear Information System (INIS)

    A new approach that uses wet-dry-dry loading technology has been successfully demonstrated, including remote viewing and loading control. The experience gained allowed for comprehensive expeditious licensing, including transport and storage permissions. A measurements campaign of more than two years has been made on the loaded cask, confirming: heat transfer, shielding, leaktightness and fuel behavior; in May 1985 the cask was transported from the reactor DIORIT to a new Away From Reactor-AFR-storage facility in the authors premises, which by the way is the first of its kind in operation licensed. In spite of the rather complex situation with the DIORIT spent fuel and the extreme limitations in the reactor handling and loading facilities, the complete project was achieved in only 22 months; accounting for two years of monitoring and measurements, the whole project took only 3.8 years. This shows the high degree of maturity achieved in spent fuel dry storage in transport cask, and reflects the high degree of technology innovation that has been demonstrated and which can be transferred as required

  15. Gender Differences in Sorting

    DEFF Research Database (Denmark)

    Merlino, Luca Paolo; Parrotta, Pierpaolo; Pozzoli, Dario

    and causing the most productive female workers to seek better jobs in more female-friendly firms in which they can pursue small career advancements. Nonetheless, gender differences in promotion persist and are found to be similar in all firms when we focus on large career advancements. These results......In this paper, we investigate the sorting of workers in firms to understand gender gaps in labor market outcomes. Using Danish employer-employee matched data, we fiend strong evidence of glass ceilings in certain firms, especially after motherhood, preventing women from climbing the career ladder...... provide evidence of the sticky floor hypothesis, which, together with the costs associated with changing employer, generates persistent gender gaps....

  16. Teleoperated robotic sorting system

    Science.gov (United States)

    Roos, Charles E.; Sommer, Edward J.; Parrish, Robert H.; Russell, James R.

    2000-01-01

    A method and apparatus are disclosed for classifying materials utilizing a computerized touch sensitive screen or other computerized pointing device for operator identification and electronic marking of spatial coordinates of materials to be extracted. An operator positioned at a computerized touch sensitive screen views electronic images of the mixture of materials to be sorted as they are conveyed past a sensor array which transmits sequences of images of the mixture either directly or through a computer to the touch sensitive display screen. The operator manually "touches" objects displayed on the screen to be extracted from the mixture thereby registering the spatial coordinates of the objects within the computer. The computer then tracks the registered objects as they are conveyed and directs automated devices including mechanical means such as air jets, robotic arms, or other mechanical diverters to extract the registered objects.

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

    International Nuclear Information System (INIS)

    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

  18. Safety analysis report for packaging: the ORNL loop transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Nelms, H.A.; Crowley, W.K.; Just, R.A.

    1977-11-01

    An evaluation of the ORNL loop transport cask demonstrating its compliance with the regulations governing the transportation of radioactive and fissile materials is presented. A previous review of the cask is updated to demonstrate compliance with current regulations, to present current procedures, and to reflect the more recent technology.

  19. Cermet Spent Nuclear Fuel Casks and Waste Packages

    International Nuclear Information System (INIS)

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

  20. Drop test of reinforced concrete slab onto storage cask

    International Nuclear Information System (INIS)

    In this research, drop tests onto full-scale casks considering the specifications of a falling object (weight, construction, drop height, etc.) demonstrate and evaluate the integrity of casks in case a heavy object drops into the storage facilities. (J.P.N.)

  1. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... COMMISSION 10 CFR Part 72 RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc... U.S. Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by...

  2. Optimization of cask capacity for long term spent fuel storage

    International Nuclear Information System (INIS)

    Within the framework of the IAEA Subprogramme of Spent Fuel Management, a new project was conceived, focusing on issues associated with the optimization of cask/container loading (capacity) with respect to long term storage and the related integrity of fuel. An initial Consultants Meeting held in November 2002 identified and discussed principal issues regarding the optimization of cask/container assembly capacity and burnup/age capability in the design of systems for long term spent fuel storage and the related integrity of fuel. Based on resulting working materials, a Technical Meeting was held in March 2003 to obtain country-specific views from both regulators and implementers on this topic. Discussions focused on the following issues relevant to cask loading optimization: fuel integrity, retrievability, zoning, burnup credit, damaged fuel, computer code verification, life of cask components, cask maintenance, performance confirmation, and records management. Follow-on actions and meetings will be pursued to develop a TECDOC on this subject. (author)

  3. 78 FR 73456 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI.... SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage... Approved Spent Fuel Storage Casks'' to include Amendment No. 9 to Certificate of Compliance (CoC) No....

  4. 78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-26

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI... public comment period. The document proposed to amend the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the ``List of Approved Spent...

  5. Resolving sorting mechanisms into exosomes

    NARCIS (Netherlands)

    Stoorvogel, Willem

    2015-01-01

    The complexity of mechanisms driving protein sorting into exosomes is only beginning to emerge. In a paper recently published in Cell Research, Roucourt et al. report that trimming of heparan sulfate side chains of syndecans by endosomal heparanase facilitates sorting into exosomes by the formation

  6. 75 FR 41369 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD Revision 1; Withdrawal

    Science.gov (United States)

    2010-07-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS HD Revision 1... HD cask system listing within the list of approved spent fuel storage casks to include Amendment No... ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to the CoC. Amendment No....

  7. 75 FR 41404 - List of Approved Spent Fuel Storage Casks: NUHOMS®

    Science.gov (United States)

    2010-07-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS HD Revision 1... cask system listing within the list of approved spent fuel storage casks to include Amendment No. 1 to... within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to the CoC....

  8. US cask requirements and industry capability survey

    International Nuclear Information System (INIS)

    The objectives of this paper are to provide an estimate of spent fuel shipping cask requirements for reactor to away-from-reactor (AFR) storage facility shipments from the present time until late in this century and to determine and document the willingness and capability of private industry to provide required future transportation services. In order to meet this objective, the Transportation Technology Center at Sandia National Laboratories sponsored Teledyne Energy Systems to conduct a survey of US industry. Results of tasks completed to carry out the objectives are reviewed

  9. Feasibility and incentives for burnup credit in spent-fuel casks

    International Nuclear Information System (INIS)

    The spent-fuel carrying capacities of previous-generation spent-fuel shipping casks have been primarily thermal and/or shielding limited. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced considerably and cask capacities become criticality limited. Using burnup credit in the design of future casks can result in increased cask capacities as well as reduced environmental impacts and savings in time and money

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pope, R B; Diggs, J M [eds.

    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. Nuclear cask testing films misleading and misused

    International Nuclear Information System (INIS)

    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

  13. Differences of Technical Requirements Between Transportation and Storage Metal Casks

    International Nuclear Information System (INIS)

    The worldwide demand of storage facilities for spent fuels discharged from nuclear power stations is increasing to maintain sustainable operation of the nuclear power stations. The spent fuels are stored at first in the fuel pools (wet storage). When the spent fuels exceed the pool storage capacity, the fuels are transferred to the other storage facility located at reactor or away from reactor, which often adopts a dry storage technology. To use metal casks is one of the options for the dry storage facilities, and some storage facilities have already utilized large metal casks, whose original design concept were developed to transport the spent fuels from nuclear power stations to a reprocessing plant by trains, trucks or by sea-going vessels. It is widely understood that the technology of transportation casks developed up to now is able to apply to the storage casks without any significant design changes. Technical requirements on the design are discussed between the storage cask and the transportation cask to confirm of the understanding based on the assumption that the large metal cask is used for transportation and storage respectively. (author)

  14. Nondestructive evaluation of monolithic transportation casks for spent nuclear fuel

    International Nuclear Information System (INIS)

    When spent fuel from nuclear reactors must be transported by rail or truck, Federal regulations require that it be enclosed in shipping casks that satisfy a number of stringent requirements. One configuration that is under consideration for such casks consists of monolithic metal cylinders approximately 17 ft. (5 m) long, 8 ft. (2.5 m) in diameter, with 14-in. (35-cm) thick walls. The casks are to be fabricated by casting or forging with one integrally closed end. The materials being considered for this application are austenitic steel, ferritic steel, and nodular cast iron. The thick walls are needed in order to absorb most of the radiation emitted by the contents. In addition, the casks must be capable of withstanding severe transportation accidents without a breach of the cask walls that would permit the escape of any radiation. The National Bureau of Standards conducted a study to evaluate the inspectability of the casks. The study showed that current NDE technology is adequate for inspecting the casks, provided that the inspection personnel are well trained in their respective methods, and that they are experienced with the equipment and their specific techniques, and have been properly qualified in this application. The capabilities of many NDE techniques were evaluated in the study. These techniques were based upon all of the principal NDE methods in use today, including ultrasonics, acoustic emission, radiology, liquid penetrants, magnetic particles, eddy currents, and visual inspection

  15. Decontamination of transport casks and of spent fuel storage facilities

    International Nuclear Information System (INIS)

    The present document provides an analysis of the technical papers presented at the meeting as well as a summary of the panel discussion. Conclusions and Recommendations: The meeting agreed that the primary source of contamination of transport casks is the production of radioactive isotopes in nuclear fuel and activation products of fuel components in nuclear reactors. The type, amount of mechanism for the release of these isotopes depend on the reactor type and fuel handling process. The widespread use of pools for the storage and handling of fuel provides an easy path for the transfer of contamination. Control of pool water conditions is essential for limiting the spread of contamination. For plants where casks are immersed in pools for loading, the immersion times should be minimised. Casks should be designed for ease of decontamination. The meeting discussed the use of stainless steel and suitable paints for coating casks. Designers should consider the appropriate coating for specific applications. The use of pressurized water for decontamination is recommended whenever possible. A number of commercially available reagents exist for decontaminating cask external surfaces. More work, however, is needed to cope with Pressurized Water Reactor crud within casks. Leaking fuel should be identified and isolated before storage in pools. Basic studies of the uptake and release of contamination from cask surfaces should be initiated. Standardization of methods of contamination measurement and instrumentation should be instituted. Refs, figs and tabs

  16. DOE procurement activities for spent fuel shipping casks

    International Nuclear Information System (INIS)

    This paper discusses the DOE cask development program established to satisfy the requirements of the NWPA. The program is designed to provide safe efficient casks on a timely schedule. The casks will be certified by the NRC in compliance with the 1987 amendment to NWPA. Private industry will be used to the maximum extent. DOE will encourage use of present cask technology, but will not hesitate to advance the state-of-the-art to improve efficiency in transport operations, provided that safety is not compromised. DOE will support the contractor's efforts to advance the state-of-the-art by maintaining a technical development effort that will respond to the common needs of all the contractors. DOE and the cask contractors will develop comprehensive and well integrated programs of test and analysis for cask certification. Finally, the DOE will monitor the cask development program within a system that fosters early identification of improvement opportunities as well as potential problems, and is sufficiently flexible to respond quickly yet rationally to assure a fully successful program

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

    International Nuclear Information System (INIS)

    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

  18. Robotic radiation survey and analysis system for radiation waste casks

    International Nuclear Information System (INIS)

    Sandia National Laboratories (SNL) and the Hanford Engineering Development Laboratories have been involved in the development of remote systems technology concepts for handling defense high-level waste (DHLW) shipping casks at the waste repository. This effort was demonstrated the feasibility of using this technology for handling DHLW casks. These investigations have also shown that cask design can have a major effect on the feasibility of remote cask handling. Consequently, SNL has initiated a program to determine cask features necessary for robotic remote handling at the waste repository. The initial cask handling task selected for detailed investigation was the robotic radiation survey and analysis (RRSAS) task. In addition to determining the design features required for robotic cask handling, the RRSAS project contributes to the definition of techniques for random selection of swipe locations, the definition of robotic swipe parameters, force control techniques for robotic swipes, machine vision techniques for the location of objects in 3-D, repository robotic systems requirements, and repository data management system needs

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Transportation cask contamination weeping: A program leading to prevention

    International Nuclear Information System (INIS)

    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 summarizes analyses of field experience and scoping experiments, and concentrates on current modelling and experimental validation efforts. The open-quotes weepingclose quotes phenomenon associated with spent fuel transportation casks (also known as open-quote sweatingclose quotes) 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 137Cs, 134Cs, and 60Co. 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 composure 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. A DOE program is underway at Sandia National Laboratories (SNL) to determine the physical and chemical processes involved in radionuclide contamination and release on transportation cask surfaces

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Interim storage and transport casks in Switzerland. COGEMA logistics experience

    International Nuclear Information System (INIS)

    The Swiss utilities have chosen two different ways for the management of their spent fuel after initial on-site cooling: (1) reprocessing at La Hague plant (COGEMA) and Sellafield plant (BNFL); (2) interim storage at the Central Interim Storage Facility called 'Zwischenlager Wuerenlingen AG' ( ZWILAG). Following international call for tenders by Swiss utilities, COGEMA LOGISTICS has been awarded several contracts for the supply of dual-purpose transport and storage casks for the interim storage of various spent fuel assemblies. All these casks belong to the family of the TN 24 dual purpose spent fuel storage casks in operation in the USA and in Belgium as well. They offer utilities a modular solution for the interim storage of spent fuel in robust metal casks which are fully suitable for off site transports. This flexible product can be readily adapted to suit individual user needs. The Leibstadt Nuclear Power Plant (KKL) has purchased nine licensed dual-purpose TN 97L spent fuel casks (97 BWR type fuel assemblies capacity). Three of them are already in operation at ZWILAG. COGEMA LOGISTICS has also delivered a dual-purpose TN 52L spent fuel casks (52 BWR type fuel assemblies capacity) presently used for transport of spent fuel for reprocessing. The Goesgen Nuclear Power Plant (KKG) has purchased four licensed dual-purpose TN 24G spent fuel casks (37 PWR type fuel assemblies capacity). They are all in operation at ZWILAG. The Muehleberg Nuclear Power Plant (BKW/KKM) has purchased two TN 24BH spent fuel casks (69 BWR type fuel assemblies capacity). At the time of this abstract, cold trials are carried out involving the shuttle transport cask TN 9/4 procured by COGEMA LOGISTICS as well. (author)

  3. Development of dual-purpose metal cask for interim storage of spent nuclear fuel (1). Outline of cask structure

    International Nuclear Information System (INIS)

    Spent fuels discharged from nuclear power plants in Japan are planed to be reprocessed at the nuclear fuel recycle plant under construction at Rokkasho-mura. Since the amount of the spent fuels exceeds that of recycled fuel, the spent fuels have to be properly stored and maintained as recycle fuel resource until the beginning of the reprocessing. For that sake, interim storage installations are being constructed outside the nuclear power plants by 2010. The storage dry casks have been practically used as the interim storage in the nuclear power plants. From this reason, the storage system using the storage dry casks is promising as the interim storage installations away form the reactors, which are under discussion. In the interim storage facilities, the storage using the dry cask of the storage metal cask with business showings, having the function of transportation is now under discussion. By employing transportation and storage dual-purpose cask, the repack equipments can be exhausted, and the reliability of the interim storage installations can be increased. Hitachi, Ltd. has been developing the high reliable and economical transportation and storage dry metal cask. In this report, the outline of our developing transportation and storage dry cask is described. (author)

  4. On Approximability of Block Sorting

    CERN Document Server

    Narayanaswamy, N S

    2011-01-01

    Block Sorting is a well studied problem, motivated by its applications in Optical Character Recognition (OCR), and Computational Biology. Block Sorting has been shown to be NP-Hard, and two separate polynomial time 2-approximation algorithms have been designed for the problem. But questions like whether a better approximation algorithm can be designed, and whether the problem is APX-Hard have been open for quite a while now. In this work we answer the latter question by proving Block Sorting to be Max-SNP-Hard (APX-Hard). The APX-Hardness result is based on a linear reduction of Max-3SAT to Block Sorting. We also provide a new lower bound for the problem via a new parametrized problem k-Block Merging.

  5. SCANS, Shipping Cask Design Safety Analysis

    International Nuclear Information System (INIS)

    1 - Description of program or function: SCANS (Shipping Cask Analysis System) is a microcomputer-based system of computer programs and databases for evaluating safety analysis reports on spent fuel shipping casks. SCANS calculates the global response to impact loads, pressure loads, and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. Analysis options are based on regulatory cases described in the Code of Federal Regulation (1983) and Regulatory Guides published by the NRC in 1977 and 1978. The system is composed of a series of menus and input entry cask analysis, and output display programs. An analysis is performed by preparing the necessary input data and then selecting the appropriate analysis: impact, thermal (heat transfer), thermally- induced stress, or pressure-induced stress. All data are entered through input screens with descriptive data requests, and, where possible, default values are provided. Output (i.e., impact force, moment and shear time histories; impact animation; thermal/stress geometry and thermal/stress element outlines; temperature distributions as iso-contours or profiles; and temperature time histories) is displayed graphically and can also be printed. 2 - Method of solution: Impact analyses use a one-dimensional dynamic beam model. Each node in the beam model has two translational and one rotational degrees of freedom. The impact code uses an explicit time-history integration scheme in which equilibrium is formulated in terms of the global external forces and internal force resultants. This formulation allows the code to track large rigid- body motion. Thus, the oblique impact problem can be calculated from initial impact through essentially rigid-body rotation to secondary impact. Lateral pressure due to lead-slump can also be calculated. Appropriate two-dimensional finite-element meshes are automatically generated for thermal, thermal-stress, and pressure- stress analyses, based on

  6. Country report France [Operation and maintenance of spent fuel storage and transportation casks/containers

    International Nuclear Information System (INIS)

    Transportation from Electricite de France and other foreign utilities to COGEMA La Hague reprocessing plant is performed with one family of casks in the 100 ton class. The experience gained in transport cask design and operation has resulted in design of transport/storage and storage only systems. There are 6 cask types for transportation only and 10 cask types for dual purpose storage and transportation. French authorities approve each cask design. Cask vendors provide training and assistance to users as well as a transportation file containing all actions and recording inspections of the cask. Maintenance frequencies are determined according to design an experience and maintenance specifications prepared. The extent of maintenance is at three levels: inspections on arrival and departure, every 3 years or 15 transports and every 6 years or 60 transports. According to French experience the cask maintenance costs over lifetime are the same as the cost of the cask itself. (author)

  7. Development of tipping-over analysis of cask subjected to earthquake strong motion

    International Nuclear Information System (INIS)

    Since a cask is vertically oriented during loading in cask-storage, it is necessary to investigate the integrity of the cask against tipping-over during strong earthquakes. The rocking and sliding behavior of the cask during strong earthquakes can be analyzed as a dynamic vibration problem for a rigid cylinder. In this paper, in order to clarify the tipping-over characteristics of a cask during strong earthquakes, the authors applied the Distinct Element Method (DEM) to the seismic response analysis of the cask. DEM was introduced by Cundall P.A. in 1971. It is based on the use of an explicit numerical scheme. The cask was considered to be a rigid polygonal element, which satisfied the equation of motion and the law of action and reaction. They examined the applicability of this code by comparison with experimental results obtained from shaking table tests using scale model casks considering the dimension of a 100 ton class full-scale cask

  8. ALARA studies on spent fuel and waste casks

    Energy Technology Data Exchange (ETDEWEB)

    Sutherland, S.H.

    1980-04-01

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

  9. ALARA studies on spent fuel and waste casks

    International Nuclear Information System (INIS)

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

  10. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of keff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

  11. New generation legal weight spent fuel shipping cask

    International Nuclear Information System (INIS)

    GA Technologies has proposed two new spent fuel shipping casks that have a capacity four times greater than comparable existing designs. The new casks, for legal weight truck shipments, can carry four PWR or nine BWR spent fuel assemblies. They were offered in response to the recent request for proposals issued by the Office of Civilian Radioactive Waste Management (OCRWM). The RFP addressed a new generation of truck and rail shipping casks that could transport intact spent fuel assemblies from nuclear reactors to a repository or a Monitored Retrievable Storage (MRS) facility. Our primary goal has been to maximize the number of fuel elements of each fuel type that a LWT cask can carry, while ensuring that the design meets all licensing requirements

  12. Transport/storage cask TN 1300 technical description

    International Nuclear Information System (INIS)

    The TN 1300 cask - developed by Transnuklear GmbH, Hanau - serves as a cask for the dry transport and storage of spent fuel elements from 1300 MW light water reactors. The cask is classified as - Typ B(U) package - fissile class II. The application is filed with the PTB, the German competent authority. The cask has a maximum capacity of 12 PWR fuel elements (Biblis type) or 33 BWR fuel elements. The maximum heat dissipation (natural convection) amounts to about 50 kW. This corresponds to a cooling period of about 2.5 years for PWR fuel elements. The handling weight of the TN 1300 is approx. 116.5 t. (orig./HW)

  13. Handbook for structural analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    This paper described structural analysis method of radioactive material transport casks for use of a handbook of safety analysis and evaluation. Safety analysis conditions, computer codes for analyses and stress evaluation method are also involved in the handbook. (author)

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

    International Nuclear Information System (INIS)

    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

  15. Thermal test and analysis of a spent fuel storage cask

    International Nuclear Information System (INIS)

    A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

  16. What drives Greek consumer preferences for cask wine?

    DEFF Research Database (Denmark)

    Chrysochou, Polymeros; Corsi, A. M.; Krystallis Krontalis, Athanasios

    2012-01-01

    , taste, origin) as opposed to extrinsic cues (brand, price, convenience packaging). Research limitations/implications – Two main strategic directions are suggested to Greek cask wine producers: they can either maintain the current approach to the market by providing a “simple”, not particularly refined...... of the cask wine consumer is. This study aims at filling this gap. Design/methodology/approach – Based on a web-based survey, the best-worst scaling (BWS) method was applied to measure the importance of attributes that Greek consumers assign when choosing cask wine. Then, a latent class clustering...... analysis based on the importance ratings of the attributes was applied in order to segment the Greek cask wine market. Findings – The most important attributes were found to be price, quality and convenience packaging, whereas brand, grape variety and origin were found to be the least important ones. In...

  17. Operating Experience in Spent Fuel Storage Casks

    International Nuclear Information System (INIS)

    A safe storage of spent fuels has been considered as one of the inevitable tasks for TEPCO for the last few decades. In order to increase flexibility for the fuel storage measures, TEPCO has been storing spent fuels in an on-site dry storage facility at Fukushima-Daiichi Nuclear Power Station. Since 1995, more than 400 fuel assemblies have been safely store. Integrity of storage casks and fuels were carefully checked by periodical inspections, which were conducted in 2000 and 2005. The next investigation will be held within a few years in order to verify the safety conditions even after a 15-year storage. These series of inspections will give plenty of useful data for the design and operation of the Mutsu facility, which will be the first off-site interim spent fuel storage facility away from any reactor site in Japan. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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

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

    International Nuclear Information System (INIS)

    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

  20. A cask maintenance facility feasibility study

    International Nuclear Information System (INIS)

    The Oak Ridge National Laboratory (ORNL) is supporting the USDOE Office of Civilian Radioactive Waste Management (OCRWM) in developing a transportation system for spent nuclear fuel (SNF) and defense high level waste (HLW) as a part of the Federal Waste Management System (FWMS). In early 1988, a feasibility study was undertaken to design a stand-alone, green field facility for maintaining the FWMS casks. The feasibility study provided an initial layout facility design, an estimate of the construction cost, and an acquisition schedule for a Cask Maintenance Facility (CMF). The study also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrieveable Storage (MRS) facility. The data, design, and estimated cost resulting from the study have been organized for use in the total transportation system decision-making process. Most importantly, the feasibility study also provides a foundation for continuing design and planning efforts. The feasibility study was based on an assumed stand-alone green field configuration because of the flexibility this design approach provides. A stand-alone facility requires the inclusion with support functions as well as the primary process facilities thus yielding a comprehensive design evaluation and cost estimate. For example, items such as roads, security and waste processing which might be shared with an integrated or collocated facility have been fully costed in the feasibility study. Thus, while the details of the facility design might change, the overall concept used in the study can be applied to other facility configurations as planning for the total FWMS develops

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

  3. ALGORITHM FOR SORTING GROUPED DATA

    Science.gov (United States)

    Evans, J. D.

    1994-01-01

    It is often desirable to sort data sets in ascending or descending order. This becomes more difficult for grouped data, i.e., multiple sets of data, where each set of data involves several measurements or related elements. The sort becomes increasingly cumbersome when more than a few elements exist for each data set. In order to achieve an efficient sorting process, an algorithm has been devised in which the maximum most significant element is found, and then compared to each element in succession. The program was written to handle the daily temperature readings of the Voyager spacecraft, particularly those related to the special tracking requirements of Voyager 2. By reducing each data set to a single representative number, the sorting process becomes very easy. The first step in the process is to reduce the data set of width 'n' to a data set of width '1'. This is done by representing each data set by a polynomial of length 'n' based on the differences of the maximum and minimum elements. These single numbers are then sorted and converted back to obtain the original data sets. Required input data are the name of the data file to read and sort, and the starting and ending record numbers. The package includes a sample data file, containing 500 sets of data with 5 elements in each set. This program will perform a sort of the 500 data sets in 3 - 5 seconds on an IBM PC-AT with a hard disk; on a similarly equipped IBM PC-XT the time is under 10 seconds. This program is written in BASIC (specifically the Microsoft QuickBasic compiler) for interactive execution and has been implemented on the IBM PC computer series operating under PC-DOS with a central memory requirement of approximately 40K of 8 bit bytes. A hard disk is desirable for speed considerations, but is not required. This program was developed in 1986.

  4. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P.; Ribeiro, M.I.; Aparicio, P. [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  5. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  6. TRANSPORTATION CASK RECEIPT AND RETURN FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Verification tests on cask-storage method for storing spent fuel at reactor

    International Nuclear Information System (INIS)

    CRIEPI has conducted a feasibility study on spent-fuel storage and has shown that shipping and storage casks are the best method for storing less than 500 tons of spent fuel at the reactor. The cask-storage facility is composed of a storage house and casks. The cask has sealing, heat conduction, shielding and criticality-prevention functions. The storage house is used for managing casks in it. In consideration of the above functions, we confirmed the integrity of cask and spent fuel under normal conditions and in hypothetical accident conditions. (J.P.N.)

  10. Impact velocity vs. target hardness relationships for equivalent response of cask structures

    International Nuclear Information System (INIS)

    In this paper, impact velocity vs. target hardness relationships for cask structures are reviewed. The relationships are based on equivalent cask responses in terms of equal deceleration or similar cask damages. By examining several past cask or container tests as well as some analytical results, some conclusions can be drawn about the relationship between target hardness and equivalent impact velocities. This relationship clearly shows that the cask response to impact is cask-dependent and that the rigid sphere impact model results in an unconservative estimate of equivalent velocity

  11. Sorting and selection in posets

    DEFF Research Database (Denmark)

    Daskalakis, Constantinos; Karp, Richard M.; Mossel, Elchanan;

    2011-01-01

    Classical problems of sorting and searching assume an underlying linear ordering of the objects being compared. In this paper, we study these problems in the context of partially ordered sets, in which some pairs of objects are incomparable. This generalization is interesting from a combinatorial...... from two decades ago by Faigle and Turán. In particular, we present the first algorithm that sorts a width-$w$ poset of size $n$ with query complexity $O(n(w+\\log n))$ and prove that this query complexity is asymptotically optimal. We also describe a variant of Mergesort with query complexity $O......(wn\\log\\frac{n}{w})$ and total complexity $O(w^{2}n\\log\\frac{n}{w})$; an algorithm with the same query complexity was given by Faigle and Turán, but no efficient implementation of that algorithm is known. Both our sorting algorithms can be applied with negligible overhead to the more general problem of reconstructing...

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

  14. Hybrid optical and acoustic force based sorting

    Science.gov (United States)

    O'Mahoney, Paul; Brodie, Graham W.; Wang, Han; Demore, Christine E. M.; Cochran, Sandy; Spalding, Gabriel C.; MacDonald, Michael P.

    2014-09-01

    We report the combined use of optical sorting and acoustic levitation to give particle sorting. Differing sizes of microparticles are sorted optically both with and without the aid of acoustic levitation, and the results compared to show that the use of acoustic trapping can increase sorting efficiency. The use of a transparent ultrasonic transducer is also shown to streamline the integration of optics and acoustics. We also demonstrate the balance of optical radiation pressure and acoustic levitation to achieve vertical sorting.

  15. A Heapify Based Parallel Sorting Algorithm

    Directory of Open Access Journals (Sweden)

    M. A.A.A. Hija

    2008-01-01

    Full Text Available Quick sort is a sorting algorithm whose worst case running time is θ(n2 on an input array of n numbers. It is the best practical for sorting because it has the advantage of sorting in place. Problem statement: Behavior of quick sort is complex, we proposed in-place 2m threads parallel heap sort algorithm which had advantage in sorting in place and had better performance than classical sequential quick sort in running time. Approach: The algorithm consisted of several stages, in first stage; it splits input data into two partitions, next stages it did the same partitioning for prior stage which had been spitted until 2 m partitions was reached equal to the number of available processors, finally it used heap sort to sort respectively ordered of non internally sorted partitions in parallel. Results: Results showed the speed of algorithm about double speed of classical Quick sort for a large input size. The number of comparisons needed was reduced significantly. Conclusion: In this study we had been proposed a sorting algorithm that uses less number of comparisons with respect to original quick sort that in turn requires less running time to sort the same input data.

  16. The Dry-Cap spent fuel storage/transport cask

    International Nuclear Information System (INIS)

    Increasing inventories of spent fuel and decreasing storage capacities at reactors are prompting development of alternative storage technologies. In the United States of America, the Department of Energy is engaged in the development of a geological repository and is committed to begin accepting fuel for permanent storage by 31 January 1998. Until this time, US utilities have assumed the responsibility for handling this material. The storage situation is also recognized in Japan and several utilities are now engaged in the development of alternative storage options. In recognition of these situations, Combustion Engineering, Inc. and Sumitomo Heavy Industries Ltd are engaged in a programme to develop and manufacture a cask capable of safety storing and transporting spent nuclear fuel. The cask is designed in accordance with US 10CFR71 and 10CFR72 criteria and has one of the largest capacities of spent fuel casks, with the ability to hold 24 PWR or 60 BWR spent fuel bundles and remain under the 125 t crane capacity of most power plants. The Dry-Cap spent fuel storage cask consists of a 16.5 ft. (5 m) long by 7.5 ft (2.27 m) diameter thick-walled steel cylinder surrounded by shielding material. Dry-Cap is a relatively simple design, easily manufactured and, unlike other cask designs, requires no external fins for cooling. Dissipation of decay heat is accomplished by natural convection between the fuel and its helium environment and the cask and its surrounding environment. One of the most important features of the Dry-Cap design is that it does not require poison material for criticality control, since the basket design utilizes credit for burnup. Taking credit for the known irradiation heating of discharged fuel, and the fact that it has a low residual reactivity, can simplify and minimize the maintenance and monitoring requirements for long term storage. The Dry-Cap cask is designed to fulfil the long and short term storage needs for utilities. (author)

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

    International Nuclear Information System (INIS)

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

  18. An economic evaluation of a storage system for casks with burnup credit

    International Nuclear Information System (INIS)

    It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

  19. Deformability-based capsule sorting

    Science.gov (United States)

    Le Goff, Anne; Munier, Nadege; Maire, Pauline; Edwards-Levy, Florence; Salsac, Anne-Virginie

    2015-11-01

    Many microfluidic devices have been developed for cancer diagnosis applications, most of which relying on costly antibodies. Since some cancer cells display abnormal mechanical properties, new sorting tools based on mechanical sensing are of particular interest. We present a simple, passive pinched flow microfluidic system for capsule sorting. The device consists of a straight microchannel containing a cylindrical obstacle. Thanks to a flow-focusing module placed at the channel entrance, capsules arrive well-centered in the vicinity of the obstacle. Pure size-sorting can be achieved at low shear rate. When increasing the shear rate, capsules are deformed in the narrow space between the pillar and the wall. The softer the capsule, the more tightly it wraps around the obstacle. After the obstacle, streamlines diverge, allowing for the separation between soft capsules, that follow central streamlines, and stiff capsules, that drift away from the obstacle with a wider angle. This proves that we have developed a flexible multipurpose sorting microsystem based on a simple design.

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

    International Nuclear Information System (INIS)

    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

  1. Vestibule and Cask Preparation Mechanical Handling Calculation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Final version dry cask storage study

    International Nuclear Information System (INIS)

    This report was prepared in response to Section 5064 of the Nuclear Waste Policy Amendments Act of 1987 (the Amendments Act--Public Law 100-203), which directs the Secretary of Energy to conduct a study of the use of dry-cask-storage technology for storing spent fuel at the sites of civilian nuclear reactors until a geologic repository is available. In conducting this study, whose results are being reported to the Congress, the Secretary was to consider such factors as costs, effects on human health and the environment, and the extent to which the Nuclear Waste Fund can and should be used to provide funds for at-reactor storage. In addition, the Secretary was to consult with the Nuclear Regulatory Commission (NRC), include NRC comments in the report, and solicit the views of State and local governments and the public. The study performed in response to these requirements was based largely on data published by the DOE or the NRC or included in documents issued by the DOE. Among the DOE documents are the 1987 MRS proposal to the Congress and a subsequent report, prepared to supply the Congress with additional information on the MRS facility. Because in evaluating dry storage at reactor sites it is necessary to take into account other options for meeting storage needs, this study covered all forms of at-reactor storage. 107 refs., 15 figs., 10 tabs

  3. Initial version, dry cask storage study

    International Nuclear Information System (INIS)

    This report was prepared to study the use of dry cask storage for storing spent fuel at the sites of civilian nuclear reactors until a geologic repository is available. In conducting this study, whose results are to be reported to the Congress, the Secretary is to consider such factors as costs, effects on human health and the environment, and the extent to which the Nuclear Waste Fund can and should be used to provide funds for at-reactor storage. In addition, the Secretary is to consult with the Nuclear Regulatory Commission (NRC), include NRC comments in the report, and solicit the views of State and local governments and the public. The study performed in response to these requirements was based largely on data published by the NRC or included in documents issued by the DOE. Among the DOE documents are the MRS proposal to the Congress and a subsequent report, prepared to supply the Congress with additional information on the MRS facility. Because in evaluating dry storage at reactor sites it is necessary to take into account other options for meeting storage needs, this study covered all forms of at-reactor storage. 85 refs., 5 figs, 12 tabs

  4. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

    Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1/2 truck cask, which has been relicensed to carry pressurized water reactor (PWR) fuel with 56,000 MWd/ton U burnup at 450 days of cooling time. New cask designs should consider the effects of high burnup for future use, even though it is not commercially desirable to include currently unneeded capability. In conclusion, the increased heat and gamma radiation of high-burnup fuels can be accommodated by additional cooling time, but the increased neutron radiation source cannot be accommodated unless the balance of neutron and gamma contributions to the overall dose rate is properly chosen in the initial cask design. Criticality control of high-burnup fuels is possible with heavily poisoned baskets, but burnup credit in licensing is a much more direct means of demonstrating criticality safety

  5. Safety Tests of Concrete Storage Cask for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    In preparation for the timely installation of interim storage facility for spent nuclear fuel (SF), KORAD is developing domestic models of SF storage systems and the concrete storage cask is one of them. A concrete cask consists of a metallic canister which confines SF with welded closure and a concrete overpack which provides radiation shielding and physical protection to the canister. The safety requirements for a SF storage cask is well established in US and summarized in regulatory guides such as NUREG-1536. KAERI has been performing tests of the concrete cask to demonstrate its safety and compliance to the regulatory requirements with high priority stipulated in NUREG-1536. The test program includes the structural performance tests under tip-over and earthquake and decay heat removal test under normal, off-normal and accident conditions. In this paper, brief introduction to the structural tests and their results are provided. Safety tests to demonstrate the safety of KORAD21C concrete storage cask were successfully performed. The structural integrity during tip-over and earthquake were demonstrated with scale model tests and the results are analyzed in comparison with safety analysis results

  6. Response of spent fuel transportation casks to explosive loadings

    International Nuclear Information System (INIS)

    Casks for the transportation of spent power reactor fuel can be exposed to explosive loadings from several causes. Exposure can come from an accident involving a propane or other hydrocarbon tanker, from an accident involving military or industrial explosives, or from deliberate sabotage. The regulations for the design of these casks do not specifically include requirements for resistance to blast loadings, but the hypothetical accident sequence that the casks are required to survive assure some measure of blast resistance. To perform accurate risk and security assessments, this blast resistance must be quantified. This paper will discuss the methodology used to determine the blast resistance of a representative rail and a representative truck spent fuel transportation cask. The methodology discussed in this paper can be used to determine the response to explosive loadings other than the one discussed in this paper or to determine the effect of explosive loadings on other casks. Due to the sensitive nature of this topic, this paper is intentionally vague on a number of parameters used in the analyses

  7. Shielding benchmark calculations of selected spent fuel storage cask experiments

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.; Tang, J.S.; Parks, C.V. (Oak Ridge National Lab., TN (United States)); Taniuchi, H. (Kobe Steel Ltd. (Japan))

    1993-01-01

    This paper describes the application of the three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of benchmark spent fuel storage cask measurements performed at the Idaho National Engineering Laboratory. A total of five storage cask problems were analyzed to determine the expected accuracies of computational analyses using well-established Monte Carlo codes. The results presented herein represent the current status of the work. Predicted neutron dose results generally compare very favorably (within 30%) with the measurements for the cask lid, bottom, and along the cask side. Gamma-ray dose rates exhibit differing trends, depending on the measurement location. For lid and bottom doses, as well as side doses near the endfittings, agreement is again within 30%, although several exceptions are seen. However, for gamma doses along the cask side and adjacent to the active fuel, a factor of 2 overprediction is noted. Investigations into the cause of these discrepancies are currently in progress.

  8. Shielding benchmark calculations of selected spent fuel storage cask experiments

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.; Tang, J.S.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Taniuchi, H. [Kobe Steel Ltd. (Japan)

    1993-03-01

    This paper describes the application of the three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of benchmark spent fuel storage cask measurements performed at the Idaho National Engineering Laboratory. A total of five storage cask problems were analyzed to determine the expected accuracies of computational analyses using well-established Monte Carlo codes. The results presented herein represent the current status of the work. Predicted neutron dose results generally compare very favorably (within 30%) with the measurements for the cask lid, bottom, and along the cask side. Gamma-ray dose rates exhibit differing trends, depending on the measurement location. For lid and bottom doses, as well as side doses near the endfittings, agreement is again within 30%, although several exceptions are seen. However, for gamma doses along the cask side and adjacent to the active fuel, a factor of 2 overprediction is noted. Investigations into the cause of these discrepancies are currently in progress.

  9. Estimated risk contribution for dry spent fuel storage cask

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) is pursuing means to risk-inform its regulations and programs for dry storage of spent nuclear fuel. In pursuit of this objective, the NRC will develop safety goals and probabilistic risk assessments for implementing risk-informed programs. This paper provides one example method for calculating the risk of a dry spent fuel storage cask under normal and accident conditions. The example is on the HI-STORM 100 cask at a proposed site containing four thousand such casks. The paper evaluates the risk to the public by determining the likelihood a welded stainless steel container will leak. In addition, the study addresses the risk at a site where 4,000 casks may be stored until the U.S. Department of Energy accepts the casks for placement in a repository. The methods used employ the PRODIGAL computer code to assess the probability of a faulty weld on a stainless steel-welded canister. These analyses are only the initial stages of a comprehensive risk study that the NRC is performing in support of its regulatory initiatives. (author)

  10. Shipping and storage cask data for spent nuclear fuel

    International Nuclear Information System (INIS)

    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

  11. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

    CANDU spent fuel discharged from the reactor core contains Pu, so, a special attention must be focussed into two directions: tracing for the fuel reactivity in order to prevent critical mass formation and personnel protection during the spent fuel manipulation. Shielding analyses, an essential component of the nuclear safety, take into account the difficulties occurred during the manipulation, transport and storage of spent fuel bundles, both for personnel protection and impact on the environment. The main objective here consists in estimations on radiation doses in order to reduce them under specified limit values. In order to perform the shielding calculations for the spent fuel transport cask three different codes were used: XSDOSE code and MORSE-SGC code, both incorporated in the SCALE4.4a system, and PELSHIE-3 code, respectively. As source of radiation one spent standard CANDU fuel bundle was used. All the geometrical and material data, related to the transport casks, were considered according to the shipping cask type B model, whose prototype has been realized and tested in the Institute for Nuclear Research Pitesti. The radial gamma dose rates estimated to the cask wall and in air, at different distances from the cask, are presented together with a comparison between the dose rates values obtained by all three recipes of shielding calculations. (authors)

  12. Production of casks acceptable for final storage by subsequent treatment of prefilled casks

    International Nuclear Information System (INIS)

    During the operation and the decommissioning of nuclear facilities also radioactive waste material which cannot be encompassed under the general standard waste categories arises. To transfer these types of waste material to interim/final repositories a conditioning/treatment is necessary in most cases. The acceptance conditions of the interim and final repositories require a conditioning considering the type of waste, the specific activities, and the casks to be used. A possible way of conditioning e. g. liquid waste (resins, filter aid, etc.) is to fill the waste into thick-wall casks, if necessary with additional shielding and subsequent drying res. draining. This presentation shall show the experiences and the results gained from the conditioning of these types of middle and higher activated waste. In the NPP Neckar (GKN) 14 ea. 200-I-rolling hoop drums and in the NPP Brokdorf (KBR) 83 ea. mouldings filled with granular resins were stored. 32 200-I-drums with higher activated filters, sludge, as well as mixed waste were located in shielded areas of the drum storage. (orig.)

  13. Application of FELTRAN to NEACRP TN12 shipping cask benchmark. [Shielding of spent nuclear fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Evans, A.M.; Winstanley, D.D.; Watmough, M.H. (British Nuclear Fuels plc, Risley (United Kingdom)); Gerber, R. (Salford Univ. (United Kingdom). Dept. of Pure and Applied Physics)

    1991-01-01

    British Nuclear Fuels plc and Imperial College have collaborated in developing the finite element neutron shielding design code FELTRAN to near production code status. FELTRAN solves the even parity form of the Boltzmann equation using a functional approach. The solution is found in one or two spatial dimensions using various orders of finite elements to specify the problem geometry. The angular dependence of the even parity flux is expressed using spherical harmonics. FELTRAN has been interfaced to ANISN formatted nuclear data libraries such as CASK and BUGLE. Anisotropic scattering may be specified to any order. Methods have been incorporated within the code to analyse systems with voids. FELTRAN is currently undergoing further development. The purpose of this paper is to consider the application of FELTRAN to a practical shield design problem. The OECD have adopted a benchmark experiment to measure the neutron and gamma ray radiation dose rates around a spent fuel transport flask. As part of an international collaboration the physical details of the flask design and contents have been provided to the nuclear industry. The objective is to perform an international comparison of the methods used in the analysis of cask shielding. BNFL is one of the companies involved, using the well established codes RANKERN and MCBEND. The FELTRAN calculations are performed using the same source and geometry data and equivalent angular flux expansions as for these two codes. FELTRAN is then compared with experimental and calculated results. (author).

  14. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising... Spent Fuel Storage Casks'' to include Amendment No. 8 to Certificate of Compliance (CoC) No....

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

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Conditions for spent fuel storage cask reapproval. 72.240... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.240 Conditions for spent fuel storage cask...

  16. 75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-07

    ... REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS HD... Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN), NUHOMS HD System listing within the ``List of Approved Spent Fuel Storage Casks''...

  17. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.230 Procedures for spent fuel storage cask...

  18. 75 FR 33736 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150--AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System... Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the NAC International, Inc. (NAC), MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

  19. Potential Multi-Canister Overpack (MCO) Cask Drop in the K West Basin South Loadout Pit

    International Nuclear Information System (INIS)

    This calculation note documents the probabilistic calculation of a potential drop of a multi-canister overpack (MCO) cask or MCO cask and immersion pail at the K West Basin south loadout pit. The calculations are in support of the cask loading system (CLS) subproject alignment of CLS equipment in the K West Basin south loadout pit

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  1. Influence of local regulations on TN dual purpose BWR casks

    International Nuclear Information System (INIS)

    Transnucleaire (Paris, France) and Transnuclear, Inc. (Hawthorne, New York, United Sates) have both developed Pressurized Water Reactor (PWR) spent fuel casks for storage and transport purposes. The products are supplied in Europe by Transnucleaire and in the United States by Transnuclear, Inc. Now the TN Group is working on a design for Boiling Water Reactor (BWR) spent fuel assemblies: the TN 52 L cask is designed for transport and storage is Switzerland, the TN 68 cask is designed for transport and storage in the United States. For storage purpose, national regulatory requirements have to be met: each country has specific demands and criteria. As a consequence, differences between the TN 52 L design and the TN 68 design for rather similar contents appear in several fields: the design work, the licensing process, the manufacturing and the operational life. (author)

  2. High temperature performance limit of containment system of transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Osamu; Saegusa, Toshiari [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Abiko Research Lab.

    1998-03-01

    The containment performance of a containment system using elastomer gaskets for transport casks under a high temperature and high pressure was clarified. Major results are as follows; (1) The deformation characteristics of the gaskets were represented by the compressive permanent strain rate (Dp). The temperature and time dependence was shown by Larson-Miller Parameter (LMP). (2) Generally, the high temperature performance limit is obtained by a value of LMP when the Dp value reaches 80%. However, the gaskets (FKM, VMQ, EPDM) used for real transport casks were not damaged and the containment performance was not deteriorated as a conservative condition. (3) Assuming that the service period of the gaskets for transport casks is 3 months or 1 year, the high temperature performance limit of the gasket made of fluorine rubber (FKM) is 202degC or 182degC, respectively, which includes safety margin. (author)

  3. High temperature performance limit of containment system of transport cask

    International Nuclear Information System (INIS)

    The containment performance of a containment system using elastomer gaskets for transport casks under a high temperature and high pressure was clarified. Major results are as follows; (1) The deformation characteristics of the gaskets were represented by the compressive permanent strain rate (Dp). The temperature and time dependence was shown by Larson-Miller Parameter (LMP). (2) Generally, the high temperature performance limit is obtained by a value of LMP when the Dp value reaches 80%. However, the gaskets (FKM, VMQ, EPDM) used for real transport casks were not damaged and the containment performance was not deteriorated as a conservative condition. (3) Assuming that the service period of the gaskets for transport casks is 3 months or 1 year, the high temperature performance limit of the gasket made of fluorine rubber (FKM) is 202degC or 182degC, respectively, which includes safety margin. (author)

  4. Plutonium detection in casks of compactable solid waste

    International Nuclear Information System (INIS)

    This report describes a method for determining plutonium in casks of compactable solid waste; it can be applied to amounts of plutonium varying from 2 to 200 grams. The principle of the method is the counting of the 380 keV γ photons from the plutonium 239; a correction is required if both zirconium 95 and niobium 95 are present in the cask. The maximum amount of zirconium 95 + niobium 95 which can be tolerated is 5 microcuries per gram of plutonium, and 300 microcuries per cask. Under the best conditions the accuracy of the measurement appears to be of the order of ±30 per cent, but experience has shown that the method is very useful as a guide to the recovery of the plutonium in the waste. In effect, for a batch of fifty measurements, the difference between the plutonium measured by this method and the plutonium recovered from the waste was equal to 10 per cent. (authors)

  5. Contract Report for Usage Inspection of KN-12 Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K

    2007-03-15

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse.

  6. Sorted.

    Science.gov (United States)

    Towers, S

    1997-04-01

    Each year in Accident and Emergency an increasing number of young people present with acute problems related to social drugs. These problems range from mild symptoms to life-threatening conditions, many of which can be extremely difficult and time consuming for staff to manage. It has become apparent that as with sex the experimental age for taking drugs is getting younger as youths are now far more 'streetwise' than their predecessors. This is one of the main reasons for this paper being written; it is imperative that staff are equipped with the appropriate knowledge to deal with the challenge and are educated about the problems associated with current drug trends. This potentially improves the quality of care and, in turn, good communication improves relationships. Ecstasy is once again becoming increasingly popular within mainstream clubs, as recently highlighted in the media, and with it reappear its problems. This article discusses the historical aspects of Ecstasy and aims to educate staff about its use and effects and provides health promotion advice for those who are involved in the care of people who take Ecstasy. PMID:9171546

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Thermal Evaluation of a KRI-BGM Shipping Cask

    International Nuclear Information System (INIS)

    Radioactive isotopes are used extensively in the fields of industry, medical treatment, food and agriculture. Use of radioactive isotopes is expected to increase continuously with the growth of each field. In order to safely transport radioactive isotopes from the place of manufacture to the place of use, a shipping package is required. Therefore KAERI is developing the KRI-BGM shipping cask to transport the Ir-192 bulk radioactive material, which is produced at the HANARO research reactor. The shipping package should satisfy the requirements which are prescribed in the Korea MOST Act 2001-23, IAEA Safety Standard Series No. TS-R-1, US 10 CFR Part 71 and the US 49 CFR Part 173. These regulatory classify the KRI-BGM shipping cask as a Type B package, and their regulatory guidelines state that the Type B package for transporting radioactive materials should be able to withstand a period of 30 minutes under a thermal condition of 800 .deg.. However, the polyurethane, which is to be used as the filling within the cavity of the KRIBGM shipping cask, has a very weak characteristic in a high temperature. Therefore it is difficult for the depleted uranium(hereafter DU), which is used as shielding material, to be protected under a thermal condition of 800 .deg.. Accordingly, the KRI-BGM shipping cask, which applied non-combustible polyurethane and fireproof materials as the filling, was fabricated. The thermal tests by using prototype cask have been performed to estimate the thermal integrity of the KRI-BGM shipping cask under a thermal condition of 800 .deg

  9. Swarm-Based Spatial Sorting

    CERN Document Server

    Amos, Martyn

    2008-01-01

    Purpose: To present an algorithm for spatially sorting objects into an annular structure. Design/Methodology/Approach: A swarm-based model that requires only stochastic agent behaviour coupled with a pheromone-inspired "attraction-repulsion" mechanism. Findings: The algorithm consistently generates high-quality annular structures, and is particularly powerful in situations where the initial configuration of objects is similar to those observed in nature. Research limitations/implications: Experimental evidence supports previous theoretical arguments about the nature and mechanism of spatial sorting by insects. Practical implications: The algorithm may find applications in distributed robotics. Originality/value: The model offers a powerful minimal algorithmic framework, and also sheds further light on the nature of attraction-repulsion algorithms and underlying natural processes.

  10. Lipid sorting revealed by SANS

    International Nuclear Information System (INIS)

    We have investigated the lipid sorting in a binary small unilamellar vesicle (SUV) composed of cone-shaped (1,2-dihexanoyl-sn-glycero-3-phosphocholine: DHPC) and cylinder-shaped (1,2-dipalmitoyl-sn-glycero-3-phosphocholine: DPPC) lipids. In order to reveal the lipid sorting we adopted a contrast matching technique of small angle neutron scattering (SANS), which extracts the distribution of deuterated lipids in the bilayer quantitatively. The SANS profile of deuterated SUVs at the contrast matching condition showed a characteristic scattering profile, indicating an asymmetric distribution of cone-shaped lipids in the bilayer. The fitting of the observed SANS profile revealed that most DHPC molecules are localized in the outer leaflet, which supports that the shape of the lipid is strongly coupled with the membrane curvature. We compared the obtained asymmetric distribution of the cone-shaped lipids in the bilayer with the theoretical prediction based on the curvature energy model. (author)

  11. Single beam atom sorting machine

    International Nuclear Information System (INIS)

    We create two overlapping one-dimensional optical lattices using a single laser beam, a spatial light modulator and a high numerical aperture lens. These lattices have the potential to trap single atoms, and using the dynamic capabilities of the spatial light modulator may shift and sort atoms to a minimum atom-atom separation of 1.52 μm. We show how a simple feedback circuit can compensate for the spatial light modulator's intensity modulation

  12. Swarm-Based Spatial Sorting

    OpenAIRE

    Amos, Martyn; Don, Oliver

    2008-01-01

    Purpose: To present an algorithm for spatially sorting objects into an annular structure. Design/Methodology/Approach: A swarm-based model that requires only stochastic agent behaviour coupled with a pheromone-inspired "attraction-repulsion" mechanism. Findings: The algorithm consistently generates high-quality annular structures, and is particularly powerful in situations where the initial configuration of objects is similar to those observed in nature. Research limitations/implications: Exp...

  13. Optimization of cask capacity for long term spent fuel storage

    International Nuclear Information System (INIS)

    Full text: Long term storage of spent fuel is a priority topic within the Member States of the IAEA. Long term spent fuel storage was previously addressed in an IAEA Co-ordinated Research Project /1/, which recognized the growing challenge of extending the life of storage facilities. Dry cask storage of spent fuel is playing a steadily increasing role in this regard. Storage practices should comply with IAEA safety requirements 'International Basic Safety Standards for Protection Against Ionizing Radiation and for the Safety of Radiation Sources' /2/, including maintaining doses as low as reasonably (taking economic/social/etc aspects. into account) achievable [i.e., the ALARA principle]. Within the framework of the IAEA Subprogramme of Spent Fuel Management, a new project was conceived, focusing on issues associated with the optimisation of cask/container loading (capacity) with respect to long term storage and the related integrity of fuel, see IAEA /3. Optimization is a part of the design process in which the combination of application objectives, regulatory limits and design margins are innovatively addressed and judiciously balanced in the final design. A primary result of a successful design optimization is a cask of superior assembly and burnup/age capacity that minimizes the total number of required cask loadings. An equally important and parallel benefit is that this process also results in reduced radiation exposure, thereby contributing significantly to maintaining doses as low as reasonably achievable (ALARA objectives). In this sense, both cask designers and regulators have the common ultimate goal of improving cask performance, and thus facilitating optimization. An initial Consultants Meeting held in November 2002 identified and discussed principal issues regarding the optimization of cask/container assembly capacity and burnup/age capability in the design of systems for long term spent fuel storage and the related integrity of fuel. Working

  14. Development of new type concrete for spent fuel storage cask

    International Nuclear Information System (INIS)

    Heat resistant concrete has been developed to make it possible to design a new type cask that has been designed on the same concept of metal cask technologies for use in high temperature conditions. The allowable temperature of conventional concrete is limited to less than 100 degrees Celsius because most of its moisture is free water and therefore hydrogen, which is effective for neutron shielding, can be easily lost. Our newly developed concrete uses chemically bonded water and as a result can be used under high temperatures

  15. Quality assurance in a cask fleet parts control system

    International Nuclear Information System (INIS)

    This paper discusses applicable portions of the eighteen Quality Assurance criteria of Subpart H, 10 CFR 71 which are incorporated into a relational data base system which has been designed to manage the spare parts control system for a fleet of spent nuclear fuel casks. The system includes not only parts in warehouse storage but parts in use in the field plus casks, ancillary equipment, test equipment, support devices, and even personnel. It provides a high degree of assurance that any device for which a condition for certification has expired will be flagged for recertification testing or removal from service well before the critical date

  16. Quality assurance in a cask fleet parts control system

    International Nuclear Information System (INIS)

    Applicable portions of the eighteen Quality Assurance criteria of Subpart H, 10 CFR 71 are incorporated into a relational data base system which has been designed to manage the spare parts control system for a fleet of spent nuclear fuel casks. The system includes not only parts in warehouse storage but parts in use in the field plus casks, ancillary equipment, test equipment, support devices, and even personnel. It provides a high degree of assurance that any device for which a condition for certification has expired will be flagged for recertification testing or removal from service well before the critical date

  17. Dry Spent Fuel Cask Transporter equipment design, testing, and operational features

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE) has established a program for the testing of a variety of dry spent fuel storage casks. The program is being conducted at the Idaho National Engineering Laboratory (INEL) by EG and G Idaho Inc. Testing of storage casks at INEL requires that large storage casks (max. gross wt. 127.1 Mg) be moved and positioned from/to an indoor loading location to an outdoor storage pad. A Dry Spent Fuel Cask Transporter has been developed to safely, conveniently, and economically transport/handle a variety of storage casks within and around the confines of nuclear sites and facility

  18. Performance testing and analyses of the VSC-17 ventilated concrete cask

    International Nuclear Information System (INIS)

    This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power ampersand Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, 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 a vertical cask orientation. Data on spent fuel integrity were also obtained

  19. Word Sorts for General Music Classes

    Science.gov (United States)

    Cardany, Audrey Berger

    2015-01-01

    Word sorts are standard practice for aiding children in acquiring skills in English language arts. When included in the general music classroom, word sorts may aid students in acquiring a working knowledge of music vocabulary. The author shares a word sort activity drawn from vocabulary in John Lithgow's children's book "Never Play…

  20. Sorting device suitable for nuclear plants

    International Nuclear Information System (INIS)

    A device is described for handling and sorting various items and especially linen from laundry bags of a nuclear plant for separation of contaminated and non contaminated objects before washing. It includes reception means, a glove box type enclosure for sorting and exit means of sorted items. Preferentially a ventilation maintains a depression inside

  1. On the Construction of Sorted Reactive Systems

    DEFF Research Database (Denmark)

    Birkedal, Lars; Debois, Søren; Hildebrandt, Thomas

    2008-01-01

    We develop a theory of sorted bigraphical reactive systems. Every application of bigraphs in the literature has required an extension, a sorting, of pure bigraphs. In turn, every such application has required a redevelopment of the theory of pure bigraphical reactive systems for the sorting at hand...

  2. Design of Garbage Sorting Machine

    Directory of Open Access Journals (Sweden)

    Stephen K. Adzimah

    2009-01-01

    Full Text Available Problem statement: Domestic waste collection, sorting and disposal are major problems in many developing countries such as Ghana. It is an undeniable fact that the environment has been engulfed in filth. This filth comprises of the garbage and waste generated in homes, workplace and industrial setups. Most of this waste has found its way into the streets, gutters, in and around the homes, dung hills and worst of all, water bodies, many of which are sources of the drinking water treated at high costs or not treated at all. Approach: Garbage needs to be sorted into various components and each of such components like textile materials, polythene, foodstuffs, metals and glassware would then have to be handled separately at the disposal or recycling site. Such a process required a certain degree of literacy, discipline and certain basic equipment, for example separate collector bins or sorting bags. In the developed world this is not much a problem because every home has different polythene bags into which the various constituents of domestic waste are put right at the generation point. Separate collection bins were also provided at vantage points for the various types of domestic garbage collection. In the developing countries these arrangements have not been feasible because of the level of literacy, lack of appreciation of the problem, non-availability of the different types of polythene bags and poverty. Currently, most garbage collection in the developing countries is done by depositing every thing into a single container from where they are hauled to be dumped in landfills or burned in incinerators. Refuse disposal by land filling requires a sizeable land for the sole purpose of refuse disposal. This may lead to (1: Encumbering large tracks of prime land, which could not be put to other uses (2: Pollution of ground water by the leachate from the landfills (3: Breeding of leaches, rodents, mosquitoes and (4: Generation of strong stench coming

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

    International Nuclear Information System (INIS)

    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 fuer 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 -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 negligible. (author)

  4. Spent fuel storage cask testing and operational experience at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Spent-fuel storage cask research, development, and demonstration activities are being performed for the U.S. Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) as a part of the storage cask testing program. The cask testing program at federal sites and other locations supports the Nuclear Waste Policy Act (NWPA) and DOE objectives for cooperative demonstrations with the cask vendors and utilities for development of at-reactor dry cask storage capabilities for spent nuclear fuel assemblies. One research and development program for the storage cask performance testing of metal storage cask was initiated through a cooperative agreement between Virginia Power and DOE in 1984. The performance testing was conducted for the DOE and the Electric Power Research Institute by the Pacific Northwest laboratory, operated for DOE by Battelle Memorial Institute, and the Idaho National Engineering Laboratory (INEL), operated for DOE by EG ampersand G Idaho, Inc. In 1988 a cooperative agreement was entered into by DOE with Pacific Sierra Nuclear Associates (PSN) for performance testing of the PSN concrete Ventilated Storage Cask. Another closely related activity involving INEL is a transportable storage cask project identified as the Nuclear Fuel Services Spent-Fuel Shipping/Storage Cask Demonstration Project. The purpose of this project is to demonstrate the feasibility of packing, transporting, and storing commercial spent fuel in dual-purpose transport/storage casks

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

    International Nuclear Information System (INIS)

    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

  6. Computational fluid dynamics analysis for K24B cask design with burnup credit

    International Nuclear Information System (INIS)

    Korea Nuclear Engineering Service Corp. (KONES) has designed K24B cask for the storage and the transportation of 24 (CE-type 16x16) PWR assemblies. K24B cask is designed with considering burnup credit of spent fuel. In order to remove heat from the fuel assemblies effectively, the flow channels in the upper and the lower part of fuel assemblies are set up to promote the natural convection. Computational fluid dynamics analysis is carried out to estimate and assure the thermal integrity of K24B cask. Conduction and radiation heat transfer through the cask components and the natural convective heat transfer in the cask are simulated. As a result of the analysis, the maximum temperatures of the cask components are maintained below the operating temperature for the safety. Therefore, the design of K24B cask can satisfy the safety limit. (author)

  7. Asymmetric temperature profiles in transport and storage casks for radioactive materials

    International Nuclear Information System (INIS)

    Transport and storage casks for spent fuel elements or vitrified radioactive waste are exposed to radioactive radiation and additional thermal load due to the radioactive inventory. A reliable heat removal is required in order to avoid material degradation of the shielding and the cask. The calculation procedures of maximum temperatures in the casks need structural modeling of the cask inventory and the environmental conditions with respect to the heat removal. The authors show that simplified models with homogeneous heat load distributions underestimate the real conditions. Detailed models of the cask internals and the air circulation around the cask under a protection cover and sun irradiation have to be taken into account. The calculations methods have to be adapted to the safety relevant conditions of each cask type.

  8. The interim storage facility with dry storage casks and its safeguards activity

    International Nuclear Information System (INIS)

    Recyclable-Fuel Storage Company (RFS) is constructing an interim storage facility of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel from nuclear power plants and to serve for about 50 years in RFSC. Metallic dry casks have already been used for dry cask storage facility at Tokai No.2 power station of Japan Atomic Power Company. But, RFSC is not exactly the same as the dry cask storage facility at Tokai No.2 power station, for example, cask transportation between facilities and no hot cells. Therefore, additional safeguards activities are necessary. The outline of the design and handling of metallic dry casks at RFSC and the currently developing status of safeguards activity such as containment and surveillance for the cask receipt and storage at RFSC, etc are described. (author)

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

    CERN Document Server

    Poulson, D; Guardincerri, E; Morris, C L; Bacon, J D; Plaud-Ramos, K; Morley, D; Hecht, A

    2016-01-01

    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 paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon 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, the cask contents can be confirmed with high confidence in less than two days exposure. Similar results can be obtained by moving a smaller detector to view the cask from multiple angles.

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

    International Nuclear Information System (INIS)

    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

  11. AUTOCASK (AUTOmatic Generation of 3-D CASK models). A microcomputer based system for shipping cask design review analysis

    International Nuclear Information System (INIS)

    AUTOCASK (AUTOmatic Generation of 3-D CASK models) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for the structural analysis of shipping casks for radioactive material. Model specification is performed on the microcomputer, and the analyses are performed on an engineering workstation or mainframe computer. AUTOCASK is based on 80386/80486 compatible microcomputers. The system is composed of a series of menus, input programs, display programs, a mesh generation program, and archive programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests

  12. How does the Shift-insertion sort behave when the sorting elements follow a Normal distribution?

    CERN Document Server

    Pal, Mita; Mahanti, N C

    2012-01-01

    The present paper examines the behavior of Shift-insertion sort (insertion sort with shifting) for normal distribution inputs and is in continuation of our earlier work on this new algorithm for discrete distribution inputs, namely, negative binomial. Shift insertion sort is found more sensitive for main effects but not for all interaction effects compared to conventional insertion sort.

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

    International Nuclear Information System (INIS)

    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

  14. Logistics management for storing multiple cask plug and remote handling systems in ITER

    International Nuclear Information System (INIS)

    Highlights: ► We model the logistics management problem in ITER, taking into account casks of multiple typologies. ► We propose a method to determine the best position of the casks inside a given storage area. ► Our method obtains the sequence of operations required to retrieve or store an arbitrary cask, given its storage place. ► We illustrate our method with simulation results in an example scenario. -- Abstract: During operation, maintenance inside the reactor building at ITER (International Thermonuclear Experimental Reactor) has to be performed by remote handling, due to the presence of activated materials. Maintenance operations involve the transportation and storage of large, heavyweight casks from and to the tokamak building. The transportation is carried out by autonomous vehicles that lift and move beneath these casks. The storage of these casks face several challenges, since (1) the cask storage area is limited in space, and (2) all casks have to be accessible for transportation by the vehicles. In particular, casks in the storage area may block other casks, so that the former has to be moved to a temporary position to give way to the latter. This paper addresses the challenge of managing the logistics of cask storage, where casks may have different typologies. In particular, we propose an approach to (1) determine the best position of the casks inside the storage area, and to (2) obtain the sequence of operations required to retrieve and store an arbitrary cask from/to a given storage place. A combinatorial optimization approach is used to obtain solutions to both these problems. Simulation results illustrate the application of the proposed method to a simple scenario

  15. Impact analysis of spent fuel dry casks under accidental drop scenarios

    International Nuclear Information System (INIS)

    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. (author)

  16. Experience Gained From Skoda VPVR/M Casks Use For RRRFR Program

    International Nuclear Information System (INIS)

    Full text: The aim of the paper is to present the Nuclear Research Institute Rez plc, Czech Republic (NRI) cask owner experience gained from the use of unique high capacity transport and storage SKODA VPVR/M cask technology. Cask is licensed for nearly all types of Russian origin research reactor fuel (package design license, transport permission). NRI is participating in the Russian Research Reactor Fuel Return (RRRFR) program incorporated in the Russian Federation - United States common activity Global Threat Reduction Initiative (GTRI) supported by IAEA. Within the scope of this project, the high enriched uranium (HEU) and low enriched uranium (LEU) spent nuclear fuel (SNF) from NRI was returned back to RF Mayak facility in November 2007 using 16 of these casks (total 549 fuel assemblies and hermetic canisters with SNF, only one combined road and rail secure shipment needed in total). Now NRI is supporting the SNF shipments from further countries research reactors - cask Users (in 2008 from Hungary and Bulgaria already completed, in 2009 from Ukraine and Poland in progress, in the following years from Serbia, Byelorussia in preparation, and next coming). The NRI role and activity in the program is based on performing the transport packaging system inspections and maintenance, assuring the transportation of the empty SKODA VPVR/M casks in special ISO containers and cask handling auxiliary equipment / accessories to the user, providing instruction and recommendations for the research reactor facility modifications, foreign country research reactor staff education and training to handle with the casks, reviewing the user cask handling operational procedures, performing special technical support (cask drying and He-leak testing), supervising the SNF loading into the casks, and loaded casks mounting into special ISO containers. These activities are covered by a long term Cask Custodian Contract between the NRI and US DOE valid until the year 2014 and Service

  17. Polarized sorting and trafficking in epithelial cells

    Institute of Scientific and Technical Information of China (English)

    Xinwang Cao; Michal A Surma; Kai Simons

    2012-01-01

    The polarized distribution of proteins and lipids at the surface membrane of epithelial cells results in the formation of an apical and a basolateral domain,which are separated by tight junctions.The generation and maintenance of epithelial polarity require elaborate mechanisms that guarantee correct sorting and vectorial delivery of cargo molecules.This dynamic process involves the interaction of sorting signals with sorting machineries and the formation of transport carriers.Here we review the recent advances in the field of polarized sorting in epithelial cells.We especially highlight the role of lipid rafts in apical sorting.

  18. Studies and research concerning BNFP: advanced cask handling studies

    International Nuclear Information System (INIS)

    Cask turnaround times at loading and unloading sites can be improved by providing better working conditions, improved safety, reduced decontamination time, training, and where practical to do so, improved facility design. This report consists of treatments of several of these topics with the common goal of improving operational efficiency

  19. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

  20. Interim Dry Storage of Spent Fuel in Casks

    International Nuclear Information System (INIS)

    French option for the back end of the fuel cycle is reprocessing of used fuel and recycling the fissile material, except some very specific fuel stored in vaults (dry conditions). Used fuel management solutions studied by AREVA for various countries allow for either direct transport to the reprocessing plant, or interim storage and transport after storage of used fuel. Interim storage solutions are wet storage or dry storage (DSC, metal casks or vault systems). When the decision on used fuel management has been postponed, some extension of interim storage duration is considered, therefore it becomes necessary to study used fuel and cask material behaviour and deterioration mechanisms. One objective of this R&D was to review research efforts on spent fuel behaviour and Dry storage experience in casks. Particularly we were interested in the assessment of retrievability of fuel after storage for further use. A review therefore, was made of the effect of storage time/ temperatures and of loading/ drying operation on used fuel integrity. R&D programmes were also carried out on the evaluation of cask materials in long term, especially materials susceptible to degradation

  1. Structural analysis of closure bolts for shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Mok, G.C.; Fischer, L.E.

    1993-04-01

    This paper identifies the active forces and moments in a closure bolt of a shipping cask. It examines the interactions of these forces/moments and suggest simplified methods for their analysis. The paper also evaluates the role that the forces and moments play in the structure integrity of the closure bolt and recommends stress limits and desirable practices to ensure its integrity.

  2. Monitored Retrievable Storage conceptual system study: cask-in-trench

    International Nuclear Information System (INIS)

    This report provides a description of the Cask-in-Trench Storage Concept which meets a specified set of requirements; an estimate of the costs of construction, operation and decommissioning of the concept; the costs required to expand the facility throughput and storage capability; and the life cycle costs of the facility. 22 figures, 34 tables

  3. Implementation of response function concept for spent fuel cask analyses

    International Nuclear Information System (INIS)

    Due to the uncertain schedule about the first disposal of the large quantity of spent nuclear fuel (SNF) accumulated at the US commercial nuclear power plants, and due to the wide range of burnups and cooling times of the SNF, it is urgent to develop a quick and realistic method for analyzing an interim-storage or shipping package of SNF. The existing method uses design-basis SNF, and it is unnecessarily conservative and therefore uneconomic. This paper demonstrates the use of response-function concept for shielding and criticality analysis for a commercial SNF shipping cask. A PC-based computer code is written for this purpose. The program allows users to perform accurate shielding and criticality analyses for any selected cask payload on real-time basis. The results are less conservative, but more realistic than that of the design-basis analyses. One must be noted, however, that the response function is cask-specific. Therefore, the concept is most beneficial to the major cask type which is to be repeatedly used for a large number of SNF shipments

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

    Science.gov (United States)

    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.

  5. Monte Carlo shipping cask calculations using an automated biasing procedure

    International Nuclear Information System (INIS)

    This paper describes an automated biasing procedure for Monte Carlo shipping cask calculations within the SCALE system - a modular code system for Standardized Computer Analysis for Licensing Evaluation. The SCALE system was conceived and funded by the US Nuclear Regulatory Commission to satisfy a strong need for performing standardized criticality, shielding, and heat transfer analyses of nuclear systems

  6. Marital Sorting and Parental Wealth

    OpenAIRE

    Kerwin Kofi Charles; Erik Hurst; Alexandra Killewald

    2011-01-01

    Using data from the Panel Study of Income Dynamics (PSID), this paper studies the degree to which spouses sort in the marriage market on the basis of parental wealth. We estimate a variety of models, including transition matrices, OLS and TSLS models to deal with measurement error in wealth reports. Our various results show that men and women in the U.S. marry spouses whose parents have wealth similar to that of their own parents; and are very unlikely to marry persons from very different par...

  7. Microfluidic systems for optical sorting

    Czech Academy of Sciences Publication Activity Database

    Ježek, Jan; Pilát, Zdeněk; Šerý, Mojmír; Kaňka, Jan; Samek, Ota; Bernatová, Silvie; Zemánek, Pavel

    Bellingham : SPIE, 2012, 86970W: 1-9. ISBN 978-0-8194-9481-8. [CPS 2012. Czech-Polish-Slovak Optical Conference on Wave and Quantum Aspects of Contemporary Optics /18./. Ostravice (CZ), 03.09.2012-07.09.2012] R&D Projects: GA MPO FR-TI1/433; GA MŠk ED0017/01/01; GA ČR GAP205/11/1687 Institutional support: RVO:68081731 Keywords : microfluidic * cell sorting * optical tweezers * Raman spectroscopy Subject RIV: BH - Optics, Masers, Lasers

  8. Design of casks: incorporating operational feedback from maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Bimet, F.; Hartenstein, M. [COGEMA Logistics, Saint Quentin (France)

    2004-07-01

    Casks are designed to conform to regulations and to client specifications. Essential areas such as easy operation, low costs of maintenance, low operation and maintenance doses, limited waste, are not explicitly covered. Notwithstanding, COGEMA LOGISTICS uses all feedback available, so that casks are designed to be easy, safe and economical to operate and maintain. Maintenance is an activity where you do spot items that old-time designers could have made better, and things that users should not have done. Thanks to quality assurance, there are a number of data available, waiting to be collected and exploited; they have to be identified, located, retrieved, and analysed. Using information such as wear, damage, use of spare parts, access problems helps to make casks ever better. It leads to more efficient concepts, and to upgrades on existing designs; it also allows to integrate environmental considerations, inter alia in the choice of materials and in maintenance methods. It is necessary for the designer to interact with the users, the cask owners, the maintenance providers, in order to gather all relevant information and events. This is made easier when all these actors are ''under one roof'', or have very close ties. This paper presents COGEMA LOGISTICS methods for collecting and analysing all these experiences waiting to be used. It explains our process for analysing data, preparing yearly reports that are made available to our designers. It describes how each new design is subject to a maintainability study, using this feedback, so that the cask safety is always assured, that radiological doses are kept to a minimum, and that operating and maintenance costs will remain as low as possible.

  9. Design of casks: incorporating operational feedback from maintenance

    International Nuclear Information System (INIS)

    Casks are designed to conform to regulations and to client specifications. Essential areas such as easy operation, low costs of maintenance, low operation and maintenance doses, limited waste, are not explicitly covered. Notwithstanding, COGEMA LOGISTICS uses all feedback available, so that casks are designed to be easy, safe and economical to operate and maintain. Maintenance is an activity where you do spot items that old-time designers could have made better, and things that users should not have done. Thanks to quality assurance, there are a number of data available, waiting to be collected and exploited; they have to be identified, located, retrieved, and analysed. Using information such as wear, damage, use of spare parts, access problems helps to make casks ever better. It leads to more efficient concepts, and to upgrades on existing designs; it also allows to integrate environmental considerations, inter alia in the choice of materials and in maintenance methods. It is necessary for the designer to interact with the users, the cask owners, the maintenance providers, in order to gather all relevant information and events. This is made easier when all these actors are ''under one roof'', or have very close ties. This paper presents COGEMA LOGISTICS methods for collecting and analysing all these experiences waiting to be used. It explains our process for analysing data, preparing yearly reports that are made available to our designers. It describes how each new design is subject to a maintainability study, using this feedback, so that the cask safety is always assured, that radiological doses are kept to a minimum, and that operating and maintenance costs will remain as low as possible

  10. Software requirements definition Shipping Cask Analysis System (SCANS)

    International Nuclear Information System (INIS)

    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

  11. Heat transfer investigations within dry spent fuel casks

    International Nuclear Information System (INIS)

    For studying the heat transfer processes and predicting the maximum spent fuel element surface temperature in a spent fuel assembly (SFA) transported in a dry cask, model experiments have been performed with a gas-filled model cask containing a simplified electrically heated model of a WWER-type SFA with 90 fuel elements. The temperature distribution of the SFA model is measured for different heat rates under vacuum in the model cask, and under normal pressure and overpressure (0.1 ... 0.7 MPa) for several cooling gases (air, argon, helium) in order to separately investigate heat transfer processes by radiation and convection/conduction. The measuring results were compared with the calculations. Computer programmes as well as simplified calculation methods for temperature prediction were developed and checked. The results obtained are also useful for thermal analyses in the field of the dry storage of SFAs in a cask or can. Specifically it was found that: The heat removal from the SFA can be considerably improved by increasing the internal cask pressure or by using helium as coolant. The radiant heat exchange in the SFA model can be calculated with sufficient accuracy by means of a computer programme developed in 1978 or by means of a simplified analytical representation shown in the final report. Both methods are directly applicable to the original SFA and useful in order to approximately calculate the maximum SFE surface temperature under normal pressure, if the fraction of heat transferred by radiation is allowed for. For the calculation of the total heat transfer a computer programme was developed and verified, which completely permits the temperature prediction of the SFA model in dependence on heat rate, type of gaseous coolant and coolant pressure. This computer programme can be directly applied to the original SFA for the calculation of the maximum SFE surface temperature

  12. Hrs and SNX3 functions in sorting and membrane invagination within multivesicular bodies.

    Directory of Open Access Journals (Sweden)

    Véronique Pons

    2008-09-01

    Full Text Available After internalization, ubiquitinated signaling receptors are delivered to early endosomes. There, they are sorted and incorporated into the intralumenal invaginations of nascent multivesicular bodies, which function as transport intermediates to late endosomes. Receptor sorting is achieved by Hrs--an adaptor--like protein that binds membrane PtdIns3P via a FYVE motif-and then by ESCRT complexes, which presumably also mediate the invagination process. Eventually, intralumenal vesicles are delivered to lysosomes, leading to the notion that EGF receptor sorting into multivesicular bodies mediates lysosomal targeting. Here, we report that Hrs is essential for lysosomal targeting but dispensable for multivesicular body biogenesis and transport to late endosomes. By contrast, we find that the PtdIns3P-binding protein SNX3 is required for multivesicular body formation, but not for EGF receptor degradation. PtdIns3P thus controls the complementary functions of Hrs and SNX3 in sorting and multivesicular body biogenesis.

  13. Sorting and Selection in Posets

    CERN Document Server

    Daskalakis, Constantinos; Mossel, Elchanan; Riesenfeld, Samantha; Verbin, Elad

    2007-01-01

    Classical problems of sorting and searching assume an underlying linear ordering of the objects being compared. In this paper, we study a more general setting, in which some pairs of objects are incomparable. This generalization is relevant in applications related to rankings in sports, college admissions, or conference submissions. It also has potential applications in biology, such as comparing the evolutionary fitness of different strains of bacteria, or understanding input-output relations among a set of metabolic reactions or the causal influences among a set of interacting genes or proteins. Our results improve and extend results from two decades ago of Faigle and Tur\\'{a}n. A measure of complexity of a partially ordered set (poset) is its width. Our algorithms obtain information about a poset by queries that compare two elements. We present an algorithm that sorts, i.e. completely identifies, a width w poset of size n and has query complexity O(wn + nlog(n)), which is within a constant factor of the in...

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

    International Nuclear Information System (INIS)

    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

  15. Issues related to the transport of a transportable storage cask after storage

    International Nuclear Information System (INIS)

    An evaluation was performed to assess whether the reliability of a transportable storage cask system and the risks associated with its use are comparable to those associated with existing transport cask systems and, if they are not, determine how the transportable storage cask system can be made as reliable as existing systems. Reliability and failure mode analyses of both transport-only casks and transportable storage casks and implementation options are compared. Current knowledge regarding the potential effects of a long-term dry storage environment on spent fuel and cask materials is reviewed. A summary assessment of the consideration for deploying a transportable storage cask (TSC) system with emphasis on preliminary design, validation and operational recommendations for TSC implementations is presented. The analyses conclude that a transportable storage cask can likely be shipped upopened by applying a combination of design considerations and operational constraints, including environmental monitoring and pretransport assessments of functional reliability of the cask. A proper mix of these constraints should yield risk parity with any existing transport cask

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

    International Nuclear Information System (INIS)

    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

  17. Fast Parallel Sorting Algorithms on GPUs

    Directory of Open Access Journals (Sweden)

    Bilal Jan

    2012-12-01

    Full Text Available This paper presents a comparative analysis of the three widely used parallel sorting algorithms: Odd-Even sort, Rank sort and Bitonic sort in terms of sorting rate, sorting time and speed-up on CPU anddifferent GPU architectures. Alongside we have implemented novel parallel algorithm: min-max butterflynetwork, for finding minimum and maximum in large data sets. All algorithms have been implementedexploiting data parallelism model, for achieving high performance, as available on multi-core GPUsusing the OpenCL specification. Our results depicts minimum speed-up19x of bitonic sort against oddevensorting technique for small queue sizes on CPU and maximum of 2300x speed-up for very largequeue sizes on Nvidia Quadro 6000 GPU architecture. Our implementation of full-butterfly networksorting results in relatively better performance than all of the three sorting techniques: bitonic, odd-evenand rank sort. For min-max butterfly network, our findings report high speed-up of Nvidia quadro 6000GPU for high data set size reaching 224 with much lower sorting time.

  18. Recyclable Waste Paper Sorting Using Template Matching

    Science.gov (United States)

    Osiur Rahman, Mohammad; Hussain, Aini; Scavino, Edgar; Hannan, M. A.; Basri, Hassan

    This paper explores the application of image processing techniques in recyclable waste paper sorting. In recycling, waste papers are segregated into various grades as they are subjected to different recycling processes. Highly sorted paper streams will facilitate high quality end products, and save processing chemicals and energy. Since 1932 to 2009, different mechanical and optical paper sorting methods have been developed to fill the demand of paper sorting. Still, in many countries including Malaysia, waste papers are sorted into different grades using manual sorting system. Due to inadequate throughput and some major drawbacks of mechanical paper sorting systems, the popularity of optical paper sorting systems is increased. Automated paper sorting systems offer significant advantages over human inspection in terms of fatigue, throughput, speed, and accuracy. This research attempts to develop a smart vision sensing system that able to separate the different grades of paper using Template Matching. For constructing template database, the RGB components of the pixel values are used to construct RGBString for template images. Finally, paper object grade is identified based on the maximum occurrence of a specific template image in the search image. The outcomes from the experiment in classification for White Paper, Old Newsprint Paper and Old Corrugated Cardboard are 96%, 92% and 96%, respectively. The remarkable achievement obtained with the method is the accurate identification and dynamic sorting of all grades of papers using simple image processing techniques.

  19. Dry cask spent fuel storage at JAPC Tokai No.2 Power Station

    International Nuclear Information System (INIS)

    The Dry cask spent fuel storage project at Tokai No.2 power station started with a geological examination of the facility site, and a design of the cask and storage facility in the mid-1990s. Considering economical efficiency and suitability for site conditions, a cask with large capacity is designed. The cask can accommodate 61 BWR fuel assemblies and is used for on-site storage only. 24 casks can be stored in the storage facility, which consists of a concrete facility building, an overhead crane and some monitoring systems. The foundation of the facility building is supported on bedrock with steel piles. Air inlets and outlets for passive natural circulation cooling are installed on the walls. The construction of the facility building and the fabrication of 7 casks began in 1999, and were completed in 2001. 8 casks for the second stage were fabricated in 2004. 2 casks for the third stage are fabricating. And 4 casks for the forth stage are in process of design. The first loaded 4 casks have been stored safely in the facility for three years since December 2001, and followed another 9 dry casks as of the end of 2007. In addition to the above spent fuel storage management at reactor site, a spent fuel storage away from reactor (AFR) is projected to start operation by 2010. We think our experiences on Tokai No.2 power station will be able to apply to the AFR storage project, such as the design of the cask and the facility. Outline of the Tokai No.2 project, experiences on the fuel loading and cask storage conditions including the monitoring data will be reported on this paper. (author)

  20. Energy efficient data sorting using standard sorting algorithms

    KAUST Repository

    Bunse, Christian

    2011-01-01

    Protecting the environment by saving energy and thus reducing carbon dioxide emissions is one of todays hottest and most challenging topics. Although the perspective for reducing energy consumption, from ecological and business perspectives is clear, from a technological point of view, the realization especially for mobile systems still falls behind expectations. Novel strategies that allow (software) systems to dynamically adapt themselves at runtime can be effectively used to reduce energy consumption. This paper presents a case study that examines the impact of using an energy management component that dynamically selects and applies the "optimal" sorting algorithm, from an energy perspective, during multi-party mobile communication. Interestingly, the results indicate that algorithmic performance is not key and that dynamically switching algorithms at runtime does have a significant impact on energy consumption. © Springer-Verlag Berlin Heidelberg 2011.

  1. Fixing the Sorting Algorithm for Android, Java and Python

    NARCIS (Netherlands)

    Gouw, C.P.T. de; Boer, F.S. de

    2015-01-01

    Tim Peters developed the Timsort hybrid sorting algorithm in 2002. TimSort was first developed for Python, a popular programming language, but later ported to Java (where it appears as java.util.Collections.sort and java.util.Arrays.sort). TimSort is today used as the default sorting algorithm in Ja

  2. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Margeanu, C.A.; Iorgulis, C. [Reactor Physics, Nuclear Fuel Performances and Nuclear Safety Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Ciocanescu, M. [Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Prava, M. [Design Department, Institute for Nuclear Research Pitesti, Campului Str, no.1, 115400 Mioveni (Romania); Margeanu, S. [Radiation Protection Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania)

    2011-07-01

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% {sup 235}U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  3. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    International Nuclear Information System (INIS)

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  4. A nonparametric Bayesian alternative to spike sorting

    OpenAIRE

    Wood, Frank; Black, Michael J.

    2008-01-01

    The analysis of extra-cellular neural recordings typically begins with careful spike sorting and all analysis of the data then rests on the correctness of the resulting spike trains. In many situations this is unproblematic as experimental and spike sorting procedures often focus on well isolated units. There is evidence in the literature, however, that errors in spike sorting can occur even with carefully collected and selected data. Additionally, chronically implanted electrodes and arrays ...

  5. Fast algorithms for sorting and searching strings

    Energy Technology Data Exchange (ETDEWEB)

    Bentley, J.L. [Bell Labs., Murray Hill, NJ (United States); Sedgewick, R. [Princeton Univ., NJ (United States)

    1997-06-01

    We present theoretical algorithms for sorting and searching multikey data, and derive from them practical C implementations for applications in which keys are character strings. The sorting algorithm blends Quicksort and radix sort; it is competitive with the best known C sort codes. The searching algorithm blends tries and binary search trees; it is faster than hashing and other commonly used search methods. The basic ideas behind the algorithms date back at least to the 1960s, but their practical utility has been overlooked. We also present extensions to more complex string problems, such as partial-match searching.

  6. Analysis technology on the thick plate free drop impact of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    The package used to transport radioactive materials, which is called by cask, must maintain the structural integrity for the requirements of hypothetical accident conditions, 9m free drop of the thick plate impact. These requirements for the cask design should be verified through test or finite element analysis to confirm the regulatory guide. In this paper, three dimensional impact analysis using ABAQUS/Explicit code under 9m free drop of the thick plate impact condition for the KSC-4 cask is performed. As the results, maximum stress intensity on each part of the cask and deformation shape of the cask is calculated and the structural intensity of the cask is evaluated by NRC Regulatory Guides. (orig.)

  7. A study on impact problems of a spent fuel shipping cask

    International Nuclear Information System (INIS)

    The design method is described of preventing a spent fuel shipping cask from being damaged in case of drop accidents. This subject was experimentally and analytically studied toward development of a new type of cask. The cask consists of shock absorbers, a cask shell structured in three layers of steel-lead-steel, internal structures and cooling water. A toroidal shell type shock absorber was employed in this case, and its characteristics were presented. The dynamic response of each structural element was also investigated. Tests were carried out using the drop tower which was constructed in conformity with IAEA regulation. A one-dimensional, lumped-mass, nonlinear spring system was used for dynamic anayses. A 1/4 scale model of the 100 ton cask which was designed on the basis of the above results was put to the test, and as a result, the validity of the design method and the structural integrity of the new cask were confirmed. (author)

  8. Size and transportation capabilities of the existing U.S. cask fleet

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A.

    2016-01-01

    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 paper describes the first applicati...

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

    Energy Technology Data Exchange (ETDEWEB)

    Muhamad, Shalina Sheik [Prototype and Plant Development Center, Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia); Hamzah, Mohd Arif Arif B. [Prototype and Plant Development Center, Technical Support Division Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia)

    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)

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

    International Nuclear Information System (INIS)

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

  12. Dry cask spent fuel storage at JAPC Tokai No.2 power station

    International Nuclear Information System (INIS)

    The Dry cask spent fuel storage project at Tokai No.2 power station started with a geological examination of the facility site, and a design of the cask and storage facility in the mid-1990s. Considering economical efficiency and suitability for site conditions, a cask with large capacity is designed. The cask can accommodate 61 BWR fuel assemblies and is used for on-site storage only. 24 casks can be stored in the storage facility, which consists of a concrete facility building, an overhead crane and some monitoring systems. The foundation of the facility building is supported on bedrock with steel piles. Air inlets and outlets for passive natural circulation cooling are installed on the walls. The construction of the facility building and the fabrication of 7 casks began in 1999, and were completed in 2001. 8 casks for the second stage were fabricated in 2004. And 6 casks for the third stage are in process of design. The first loaded 4 casks have been stored safely in the facility for three years since December 2001, and followed another 6 dry casks as of the end of 2004. In addition to the above spent fuel storage management at reactor site, a spent fuel storage away from reactor (AFR) is projected to start operation by 2010. We think our experiences on Tokai No.2 power station will be able to apply to the AFR storage project, such as the design of the cask and the facility. Outline of the Tokai No.2 project, experiences on the fuel loading and cask storage conditions including the monitoring data will be reported on this paper. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Castor transport and storage casks for VVER and RBMK fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Gartz, R.; Gobler, A.; John, R.; Diersch, R. [GNB Gesellschaft fur Nuklear-Behalter mbH, Essen (Germany); Nemec, P. [Skoda Nuclear Machinery Plzen (Czech Republic)

    1998-12-31

    CASTOR casks have been successfully developed, manufactured and delivered for Russian type reactor fuel assemblies. These casks fulfill both the requirements for type B packages according to IAEA regulations and the requirements covering different accident situations to be assumed at the storage site. In the following, the CASTOR casks CASTOR 440/84, CASTOR RBMK and CASTOR VVER 1000 are described, the nuclear content is characterized and an overview about the status of licensing, manufacturing and delivery is given. (authors) 3 refs.

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

    Science.gov (United States)

    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. Design and realization of sort manipulator of crystal-angle sort machine

    Science.gov (United States)

    Wang, Ming-shun; Chen, Shu-ping; Guan, Shou-ping; Zhang, Yao-wei

    2005-12-01

    It is a current tendency of development in automation technology to replace manpower with manipulators in working places where dangerous, harmful, heavy or repetitive work is involved. The sort manipulator is installed in a crystal-angle sort machine to take the place of manpower, and engaged in unloading and sorting work. It is the outcome of combing together mechanism, electric transmission, and pneumatic element and micro-controller control. The step motor makes the sort manipulator operate precisely. The pneumatic elements make the sort manipulator be cleverer. Micro-controller's software bestows some simple artificial intelligence on the sort manipulator, so that it can precisely repeat its unloading and sorting work. The combination of manipulator's zero position and step motor counting control puts an end to accumulating error in long time operation. A sort manipulator's design in the practice engineering has been proved to be correct and reliable.

  18. Mediatized play

    DEFF Research Database (Denmark)

    Johansen, Stine Liv

    Children’s play must nowadays be understood as a mediatized field in society and culture. Media – understood in a very broad sense - holds severe explanatory power in describing and understanding the practice of play, since play happens both with, through and inspired by media of different sorts........ In this presentation the case of ‘playing soccer’ will be outlined through its different mediated manifestations, including soccer games and programs on TV, computer games, magazines, books, YouTube videos and soccer trading cards....

  19. Bonner sphere neutron spectrometry at spent fuel casks

    CERN Document Server

    Rimpler, A

    2002-01-01

    For transport and interim storage of spent fuel elements from power reactors and vitrified highly active waste (HAW) from reprocessing, various types of casks are used. The radiation exposure of the personnel during transportation and storage of these casks is caused by mixed photon-neutron fields and, frequently, the neutron dose is predominant. In operational radiation protection, survey meters and even personal dosemeters with imperfect energy dependence of the dose-equivalent response are used, i.e. the fluence response of the devices does not match the fluence-to-dose equivalent conversion function. In order to achieve more accurate dosimetric information and to investigate the performance of dosemeters, spectrometric investigations of the neutron fields are necessary. Therefore, fluence spectra and dose rates were measured by means of a simple portable Bonner multisphere spectrometer (BSS). The paper describes briefly the experimental set-up and evaluation procedure. Measured spectra for different locat...

  20. Fuel Element Transfer Cask Modelling Using MCNP Technique

    Science.gov (United States)

    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.

  1. Safety engineering achievements in handling casks at La Hague

    International Nuclear Information System (INIS)

    Interest is focused on safety aspects of some new trends in commercial fuel reprocessing plants at La Hague. The first is the dry cask unloading, unique in size, avoiding several meters height handling and associated risks. Moreover, improvements were introduced about contamination retention, effluent decrease and contact work time, resulting in lower operators' doses. Extensive use of standard equipment, whose replacement using a special cask is foreseen as a common maintenance operation, is another major improvement for an industrial process, increasing plant availability with low personal doses compared to contact intervention. Associated crane use led to systematic studies of falling accidents and, where necessary, improved reliability crane design. It has been shown that the design and quality of corresponding elements is such that the prevention of risk is sufficient to reach a high level of safety. 1 fig

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

    International Nuclear Information System (INIS)

    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 Caskclose 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

  3. Stress analysis of closure bolts for shipping casks

    International Nuclear Information System (INIS)

    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

  4. Storage cask drop test on reinforced concrete slab

    International Nuclear Information System (INIS)

    The test results obtained may be summarized as follows: (1) The strain and acceleration during oblique dropping are sufficiently small compared with those during vertical and horizontal dropping. The strain and acceleration due to the secondary collision after dropping are also sufficiently small as compared with those due to the primary collision. For evaluation of integrity against vertical and horizontal orientation, therefore, it can be considered that dropping in the oblique orientation will pose no problem in making such evaluation. (2) The structural integrity of the cask against its dropping at the normal operating height and up to the maximum lifting height which is determined by the construction of storage facilities was verified. (3) Since the estimated critical drop height is sufficiently heigh as compared with the above-mentioned drop height, it was verified that the cask had a sufficient margin against a falling accident during operation. (J.P.N.)

  5. Safety analysis of spent fuel transport and storage casks under extreme impact conditions

    International Nuclear Information System (INIS)

    Full text: Worldwide the security of transport and storage of spent fuel with respect to terrorism threats is a matter of concern. In Germany a spent nuclear fuel management program was developed by the government including a new concept of dry on-site interim storage instead of centralized interim storage. In order to minimize transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities, the operators of NPPs have to erect and to use interim storage facilities for spent nuclear fuel on the site or in the vicinity of nuclear power plants. Up to now, 11 on-site interim storage buildings, one storage tunnel and 4 on-site interim storage areas (preliminary cask storage till the on-site interim storage building is completed) have been licensed at 12 nuclear power plant sites. Inside the interim storage buildings the casks are kept in upright position, whereas at the preliminary interim storage areas horizontal storage of the casks on concrete slabs is used and each cask is covered by concrete elements. Storage buildings and concrete elements are designed only for gamma and neutron radiation shielding reasons and as weather protection. Therefore the security of spent fuel inside a dual purpose transport and storage cask depends on the inherent safety of the cask itself. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. Since the terror attacks of 11 September 2001 the determination of casks' inherent safety also under extreme impact conditions due to terrorist attacks has been of our increasing interest. With respect to spent fuel storage one of the most critical scenarios of a terrorist attack for a cask is the centric impact of a dynamic load onto the lid-seal-system caused e.g. by direct aircraft crash or its engine as well as by a

  6. Evaluation of Equivalent Dose Rate of Interim Dry Storage Casks Loaded with Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Equivalent dose rate calculations of the CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks were performed using SCALE 4.3 computer codes system. These casks are planned for an interim storage of spent nuclear fuel at Ignalina NPP. The dose rate calculations were made on the sidelong, upper and lower surface of the cask and at the certain distance. Results show that dose rate values on the surface of the cask are much less then permissible value 1000 μSv/h when average burnup of fuel assembly is 20 GWd/tU. (author)

  7. Shielding analysis of dual purpose casks for spent nuclear fuel under normal storage conditions

    International Nuclear Information System (INIS)

    Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, UO2 pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a 2 x 10 cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the 2 x 10 cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the 2 x

  8. Safety evaluation of dry-cask storage facility for spent fuel during earthquake

    International Nuclear Information System (INIS)

    Design criteria of storage facilities were established considering the special circumstances of Japan, such as limited site area and strong earthquakes. Therefore, it is necessary to confirm the integrity in the rare case of the collapse of storage building and gantry crane, the cooling performance of cask buried in concrete rubbish, and the resistance for overture of dry storage cask during earthquake. This report evaluated the security for impact load of falling body such as concrete wall or gantry crane, the stability of the dry storage cask during earthquake, and the cooling performance of cask. (author)

  9. Beneficial uses shipping system (BUSS) cask, safety analysis report for packaging: Volumes 1 and 2

    International Nuclear Information System (INIS)

    The Beneficial Uses Shipping System (BUSS) cask Safety Analysis Report for Packaging (SARP) was originally prepared by Sandia National Laboratory (SNL). After the certification process was completed, the ownership of the BUSS cask and associated SARP was transferred from SNL to the DOE Hanford site in Richland, Washington. During timely renewal of the BUSS cask certificate of compliance, the SARP was revised to (1) respond to the timely renewal questions, (2) consolidate the previous revision made by SNL, and (3) bring the SARP into compliance with the 1996 version of 10 CFR 71. Since the BUSS cask is now the responsibility of RL, the SARP was reissued as a Hanford document

  10. STABILITY EVALUATION OF METAL CASK ATTACHED TO A TRANSFER PALLET DURING LONG-PERIOD SEISMIC MOTIONS

    Science.gov (United States)

    Kawaguchi, Shohei; Shirai, Koji; Kanazawa, Kenji

    Rocking behavior of unfixed body is affected by center of mass, material coefficient of restitution and so on. 2/5 scale metal cask model considering these parameter was used for seismic test to evaluate stability of grounding metal cask attached to a transfer pallet under the influence of long-period earthquake motion. The newest knowledge from seismic test indicates seismic motion with high velocity over 100 kine not always cause the raise of response velocity of metal cask because of energy consumption by cask sliding and impact deformation of concrete. And new estimation method (called "Window energy spectrum method") of earthquake response spectrum gives suitable evaluation of response energy.

  11. Human factors engineering applications to the cask design activities of the Civilian Radioactive Waste Management Program

    International Nuclear Information System (INIS)

    The use of human factors engineering (HFE) in the design and use of spent fuel casks being developed for the Department of Energy's Civilian Radioactive Waste Management Program is addressed. The safety functions of cask systems are presented as background for HFE considerations. Because spent fuel casks are passive safety devices they could be subject to latent system failures due to human error. It is concluded that HFE should focus on operations and verifications tests, but should begin, to the extent possible, at the beginning of cask design. Use of HFE during design could serve to eliminate or preclude opportunity for human error

  12. Experimental studies of free-standing spent fuel storage cask subjected to strong earthquakes

    International Nuclear Information System (INIS)

    Concrete cask spent fuel storage system is considered to essentially have an economical advantage and becoming widely used. For vertically free-standing concrete cask on the floor pad in the cask storage facility, its tipping-over and sliding behavior during earthquake is one of the technical key issues to guarantee its safe performance. In this paper, the experimental studies are reported by performing the excitation test with a scale model concrete cask using two-dimensional shaking table and the applicability of the energy spectrum approach is discussed. (author)

  13. Sampled control of vibration in suspended cask by using vibration manipulation functions

    International Nuclear Information System (INIS)

    Safe and reliable operation is most important for decommissioning the Fukushima 1 nuclear power plant. Especially it requires for transferring spent nuclear fuels from fuel pool to storage cask. Since the heavy cask will be suspended during the transferring operation, there is a risk of dropping it in case of the strike of large earthquakes. In this study, we introduce analytical functions to suppress residual vibration of a suspended cask by using vibration manipulation function. Hence the oscillation of the cask can be feedforward or sampled-data controlled by moving a trolley with analog actuator, the possible risk could be reduced. (author)

  14. CERCA 01: a new safe multi-design MTR transport cask

    International Nuclear Information System (INIS)

    CERCA, a subsidiary company of FRAMATOME ANP, manufactures fuel for research reactors all over the world. To comply with customer requirements, fabrication of material testing reactors elements is a mixed of various parameters. Worldwide transportation of elements requires a flexible cask, which accommodates different designs and meets international transportation regulations. To be able to deliver most of fuel elements, and to cope with non-validation of casks used previously, CERCA decided to design its own cask. All regulatory tests were successfully performed. They completely validated and qualified the safety of this new cask concept. No matter the accidental conditions are, a 5 % ΔK subcriticality margin is always met

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

    International Nuclear Information System (INIS)

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

  16. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

  17. Safety analysis report for packaging: the ORNL HFIR spent-fuel-element shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Eversole, R.E.; Just, R.A.; Llewellyn, G.H.

    1977-11-01

    The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) spent-fuel-element shipping cask is used to transport HFIR, Oak Ridge Research Reactor (ORR), and other reactor fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures and tests were used to determine behavior of the cask relative to the general standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

  18. The NINO [No Inspector, No Operator system] cask-loading safeguards system

    International Nuclear Information System (INIS)

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

  19. Experience from transport casks management in the Swedish AFR-facility CLAB

    International Nuclear Information System (INIS)

    From the 12 Swedish reactors, all located at the coast, about 250 tonnes of spent fuel are transported to the central intermediate storage facility, CLAB, yearly. The fuel assemblies are transported in TN 17/2 type B(U) casks. The sea transport system consists of the special designed ship M/S SIGYN, 4 terminal vehicles, 12 load carriers, 10 TN 17/2 casks and 2 core component casks TN 17-CC. 26 steel containers, IP-2, for low and intermediate wastes are also part of the system. During a normal year, about 95 casks are transported to and handled at CLAB. Following the ''Green book'' manual, every month 1 cask is going through a preventing maintenance and inspection programme at the maintenance workshop in CLAB. The central intermediate storage facility for spent nuclear fuel, CLAB, was taken into active operation in July 1985. Until December 1988, 362 transport casks have been handled at CLAB and 847 tonnes of spent fuel were stored in the water pools. The CLAB facility is designed in such a way that the possibility for surface contamination of the casks is very small. If this should happen there are different systems and facilities for decontamination. So far only 1 cask has been contaminated during the unloading operation in CLAB. The successful management and the experience of handling the transport casks in CLAB, will be described in the paper. (author). 6 figs

  20. Safety margins of spent fuel transport and storage casks considering aircraft crash impacts

    International Nuclear Information System (INIS)

    The safety of spent fuel transport casks in severe accident conditions is always a matter of concern. This paper surveys German missile impact tests that have been carried out in the past to demonstrate that German cask designs for transport and interim storage are safe even under conditions of an aircraft crash impact. A fire test with a cask beside an exploding propane vessel and temperature calculations concerning prolonged fires also show that the casks have reasonably good safety margins in thermal accidents beyond regulatory fire test conditions. (author)

  1. Numerical fracture analysis for the structural design of CASTOR casks

    International Nuclear Information System (INIS)

    The numerical implementation of the dynamic J-Integral is presented as one method to compute the dynamic stress intensity factor (DSIF). The applicability of the computational method is demonstrated by a finite element simulation of a free drop test of a ductile cast iron CASTOR cask with a pre-crack. The results of the simulation are contrasted with the data from the real experiment. (author)

  2. Certification of a spent fuel cask for storage and transportation

    International Nuclear Information System (INIS)

    This paper addresses the US Nuclear Regulatory Commission's requirements for the dry storage and transportation of spent fuel, focusing on how the performance standards differ between storage and transportation. The paper also discusses the NRC cask review process, and some current issues in each area of certification. In addition, some of the issues associated with the US Department of Energy's proposed multi-purpose canister are discussed

  3. Structural evaluation of spent fuel dry storage cask

    International Nuclear Information System (INIS)

    In a various regulations and standards related to the spent fuel storage, the storage casks should be designed to sustain the structural integrity under the accident conditions of predicted operation and design criteria. These conditions for the structural evaluation requires the drop, tip-over, wind like tornado and typhoon, flood and earthquake. This paper describes the load cases and conceptual evaluation method for the structural evaluation. Preliminary safety analysis of the concrete storage system were performed

  4. CLASSIFICATION OF THE MGR CARRIER/CASK TRANSPORT SYSTEM

    International Nuclear Information System (INIS)

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) carrier/cask transport system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  5. Cask operation and maintenance for spent fuel storage

    International Nuclear Information System (INIS)

    Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage

  6. Spent fuel transportation cask response to a tunnel fire scenario

    International Nuclear Information System (INIS)

    On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. 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 an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation cask designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the ANSYS registered and COBRA-SFS computer codes to evaluate the thermal performance of different cask designs. The staff concluded that the transportation casks analyzed would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event. No release of radioactive materials would result from exposure of the casks analyzed to such an event. This paper describes the methods and approach used for this assessment

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

    International Nuclear Information System (INIS)

    The development of the Source Term Analyses for Containment Evaluations (STACE) methodology provides a unique means for estimating the probability of cladding breach within transport casks, quantifying the amount of radioactive material released into the cask interior, and calculating the releasable radionuclide concentrations and corresponding maximum permissible leakage rates. Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source team has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volitile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking in which experimental validation is planned. Finally, the ANSI N14.5 recommendation that 3% and 100% of the fuel rods fail during normal and hypothetical accident conditions of transport, respectively, has been show to be overly conservative by several degrees of magnitude for these example analyses. Furthermore, the maximum permissible leakage rates for this example assembly under normal and hypothetical accident conditions are significanly higher that the leaktight requirements. By relaxing the maximum permissible leakage rates, the source term methodology is expected to significantly improvecask economics and safety

  8. Cask system maintenance in the Federal Waste Management System

    International Nuclear Information System (INIS)

    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

  9. Conceptual evaluation of metal storage cask for conditioned spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Seo, K. S.; Shin, H. S.; Lee, J. C.; Bang, K. S.; Kim, H. D.; Park, S. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-07-01

    The storage parameters of spent PWR fuel are radioactivity, heat power and its volume. Those values could be reduced to about a quarter by an Advanced spent fuel Conditioning Process (ACP). Firstly, a storage concept and scenario were established considering the characteristics of conditioned spent fuel. If the efficiency of the cooling system is improved and the appropriate quantities of the conditioned spent fuel are stored, the conditioned spent fuels could be stored at a four times higher level of spent fuel storage. One storage unit of conditioned spent fuel was designed to have its capacity equivalent to one PWR spent fuel. It was supposed that a metal storage cask has 7 baskets that can load 28 storage units. Those capacities means that 28 spent PWR fuels in metal casks can be stored. The conceptual evaluations of the critical, shielding, thermal and structural fields were performed. In conclusion, the results of the conceptual evaluations show that the proposed metal cask satisfied with the important design criteria at a smaller size than the existing systems.

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

    International Nuclear Information System (INIS)

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

  11. Performance of bolted closure joint elastomers under cask aging conditions

    Energy Technology Data Exchange (ETDEWEB)

    Verst, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    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.

  12. Engineering a Cache-Oblivious Sorting Algorithm

    DEFF Research Database (Denmark)

    Brodal, Gerth Stølting; Fagerberg, Rolf; Vinther, Kristoffer

    2007-01-01

    This paper is an algorithmic engineering study of cache-oblivious sorting. We investigate by empirical methods a number of implementation issues and parameter choices for the cache-oblivious sorting algorithm Lazy Funnelsort, and compare the final algorithm with Quicksort, the established standard...

  13. Data Sorting Using Graphics Processing Units

    Directory of Open Access Journals (Sweden)

    M. J. Mišić

    2012-06-01

    Full Text Available Graphics processing units (GPUs have been increasingly used for general-purpose computation in recent years. The GPU accelerated applications are found in both scientific and commercial domains. Sorting is considered as one of the very important operations in many applications, so its efficient implementation is essential for the overall application performance. This paper represents an effort to analyze and evaluate the implementations of the representative sorting algorithms on the graphics processing units. Three sorting algorithms (Quicksort, Merge sort, and Radix sort were evaluated on the Compute Unified Device Architecture (CUDA platform that is used to execute applications on NVIDIA graphics processing units. Algorithms were tested and evaluated using an automated test environment with input datasets of different characteristics. Finally, the results of this analysis are briefly discussed.

  14. An analysis of contingencies for making casks available for use during the early years of Federal Waste Management System operations

    International Nuclear Information System (INIS)

    A study has been performed to examine the contingencies that could be pursued by the Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) for shipping spent fuel beginning in 1998. OCRWM's current plan is to initiate operations using early production units of Initiative I truck and rail/barge casks that are presently being designed. Contingencies to this plan were considered in case some unforeseen event occurs that precludes the Initiative I casks from entering into service early in 1998 in sufficient quantities (both numbers and types) to satisfy DOE's shipping needs. Specifically, the study addressed the potential availability of cask systems, selected several cask usage scenarios, determined the requirements for casks under these scenarios, generically assessed different strategies for acquiring casks or the use of casks, and generically assessed cask fabrication capabilities. Issues concerning both domestic and foreign resources were addressed with a focus on the first five years of Federal Waste Management System (FWMS) operation

  15. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    International Nuclear Information System (INIS)

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO2, Al2O3, Gd2O3, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO2) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al2O3) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO2 for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate casks with variable cermet compositions as a function of

  16. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  17. Al analysis and design of dry storage cask of spent nuclear fuel

    International Nuclear Information System (INIS)

    According to thermal analysis of the existing concrete cask, the maximum temperature occurred at the upper side of cask. If the cask lid is made of concrete, the temperature of concrete in lid exceeds the allowable value. Based on that result, research is progressed focusing on two strategies - one is to increase thermal margin, another is to decrease the lid concrete temperature. Here, thermally - enhanced design is suggested and analyzed. This design features the air flow duct in the lid and the thermal shielding disk installed between canister and lid. Air flow duct on the center of lid concrete connected to existing air outlet can decrease temperature by promoting the convection heat transfer, and thermal shielding disk bearing smaller hole located on the center can maintain the increased convection heat transfer and minimize radiation heat transfer from canister to lid concrete for the lid concrete temperature not to be over the allowable value. Thermal analysis result for this design shows that it can be very successful to achieve these objectives. The overall component of cask temperature decrease by 2-10 .deg. C, and the lid concrete temperature dropped from above 100 to 87.5 .deg. C which is below the allowable value 93 .deg. C. In addition, heat removal of cask depending on distance between casks was investigated. Cask heat is removed by convection and radiation heat transfer at an external surface to environment. Naturally, these heat transfers are mainly affected by ambient temperature. The ambient temperature can be affected if the thermal boundary layer is overlapped. So, thermal boundary layer thickness of cask was calculated to estimate to see if the ambient temperature is affected by other cask. Boundary layer thickness is calculated is too small just about 5cm. It is concluded that distance between casks can do little impact on heat removal of cask in a practical view

  18. Enhancement of Selection, Bubble and Insertion Sorting Algorithm

    Directory of Open Access Journals (Sweden)

    Muhammad Farooq Umar

    2014-07-01

    Full Text Available In everyday life there is a large amount of data to arrange because sorting removes any ambiguities and make the data analysis and data processing very easy, efficient and provides with cost less effort. In this study a set of improved sorting algorithms are proposed which gives better performance and design idea. In this study five new sorting algorithms (Bi-directional Selection Sort, Bi-directional bubble sort, MIDBiDirectional Selection Sort, MIDBidirectional bubble sort and linear insertion sort are presented. Bi-directional Selection Sort and MIDBiDirectional Selection Sort are the enhancement on basic selection sort while Bidirectional bubble sort and MIDBidirectional bubble sort are the enhancement on basic bubble sort by changing the selection and swapping mechanism of data for sorting. Enhanced sorting algorithms reduced the iteration by half and quarter times respectively. Asymptotically complexities of these algorithms are reduced to O (n2/2 and O (n2/4 from O (n2. Linear insertion sort is the enhancement of insertion sort by changing the design of algorithm (convert two loops to one loop. So asymptotically this algorithm is converted to linear time complexity from quadratic complexity. These sorting algorithms are described using C. The proposed algorithms are analyzed using asymptotic analysis and also using machine-running time and compared with their basic sorting algorithms. In this study we also discuss how the performance and complexity can be improved by optimizing the code and design.

  19. An Unsupervised Online Spike-Sorting Framework.

    Science.gov (United States)

    Knieling, Simeon; Sridharan, Kousik S; Belardinelli, Paolo; Naros, Georgios; Weiss, Daniel; Mormann, Florian; Gharabaghi, Alireza

    2016-08-01

    Extracellular neuronal microelectrode recordings can include action potentials from multiple neurons. To separate spikes from different neurons, they can be sorted according to their shape, a procedure referred to as spike-sorting. Several algorithms have been reported to solve this task. However, when clustering outcomes are unsatisfactory, most of them are difficult to adjust to achieve the desired results. We present an online spike-sorting framework that uses feature normalization and weighting to maximize the distinctiveness between different spike shapes. Furthermore, multiple criteria are applied to either facilitate or prevent cluster fusion, thereby enabling experimenters to fine-tune the sorting process. We compare our method to established unsupervised offline (Wave_Clus (WC)) and online (OSort (OS)) algorithms by examining their performance in sorting various test datasets using two different scoring systems (AMI and the Adamos metric). Furthermore, we evaluate sorting capabilities on intra-operative recordings using established quality metrics. Compared to WC and OS, our algorithm achieved comparable or higher scores on average and produced more convincing sorting results for intra-operative datasets. Thus, the presented framework is suitable for both online and offline analysis and could substantially improve the quality of microelectrode-based data evaluation for research and clinical application. PMID:26711713

  20. 76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    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 ``List of Approved Spent Fuel Storage Casks'' as Certificate of Compliance Number 1032. DATES:...

  1. 76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI90 List of Approved Spent Fuel Storage Casks: HI.... Nuclear Regulatory Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations to add the HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage...

  2. 78 FR 22411 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections

    Science.gov (United States)

    2013-04-16

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No... direct final rule amending its spent fuel storage regulations by revising the Holtec International, Inc. (Holtec) HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks''...

  3. Overview of research and development of metal cask for transport and storage of spent nuclear fuel in Japan

    International Nuclear Information System (INIS)

    The paper overviews experimental studies of dual-purpose metal casks carried out in Japan. Full-scale casks were dropped onto a reinforced concrete target simulating hypothetical accidental drop during handling procedure in a storage facility. In some cases, leakage from the primary lid was detected, but no leakage from the secondary lid. A heavy weight drop test was carried out onto a full-scale cask simulating hypothetical collapse of a storage building due to earthquake, etc. The cask maintained its integrity. A full-scale cask was covered with a thermal insulator simulating a hypothetical burial by debris due to a building collapse in earthquake, etc. Some components might need to be recovered from the debris before reaching their critical temperature. A scale-model of a cask was subjected to seismic motion on a shaking table simulating an earthquake. The cask was rocking more for an earthquake with longer wavelength. Long-term containment of metal gaskets in double lid structure of casks has been tested with full-scale lid model. Transportability of cask after long-term storage was tested simulating degradation of cask components. Effects of aging of cask body metal, basket metal, seal and neutron shielding materials were investigated. With those degradations, cask performance in terms of shielding, sub-criticality, heat removal and containment were investigated. (author)

  4. Licensing and safety issues associated with dry cask storage update. Panel Discussion

    International Nuclear Information System (INIS)

    Full text of publication follows: Panelists from the nuclear industry, cask vendors, the U.S. Department of Energy (DOE), and the U.S. Nuclear Regulatory Commission will speak to the current status of licensing casks for interim storage and shipping to the DOE permanent site and alternate interim private storage initiatives. Subject coverage will include a broad range of relevant issues. (authors)

  5. 76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... COMMISSION 10 CFR Part 72 RIN 3150--AI89 List of Approved Spent Fuel Storage Casks: NUHOMS HD System Revision... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN) NUHOMS HD System listing within the ``List of Approved Spent Fuel...

  6. 78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks... Regulatory Commission (NRC) is proposing to amend its spent fuel storage regulations by revising the NAC... within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 3 to Certificate...

  7. 76 FR 2243 - List of Approved Spent Fuel Storage Casks: NUHOMS ® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... 3150-AI89 List of Approved Spent Fuel Storage Casks: NUHOMS HD System Revision 1 AGENCY: Nuclear... Commission) is amending its spent fuel storage regulations by revising the Transnuclear, Inc. (TN) NUHOMS HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1...

  8. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181... spent fuel storage cask designs. The NRC subsequently issued a final rule on November 21, 2008 (73 FR... COMMISSION 10 CFR Part 72 RIN 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System,......

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

    Science.gov (United States)

    2010-07-21

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150--AI88 List of Approved Spent Fuel Storage Casks: NAC.... Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by... 72. PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL,...

  10. 75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision... Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

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

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Specific requirements for spent fuel storage cask approval... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.236 Specific requirements...

  12. 76 FR 70374 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising... 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL...

  13. 77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2012-01-27

    ... 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 2 AGENCY: Nuclear... amended the NRC's spent fuel storage regulations by revising the NAC International, Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 2...

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

    International Nuclear Information System (INIS)

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

  15. Regulation of dopamine release by CASK-β modulates locomotor initiation in Drosophila melanogaster

    Directory of Open Access Journals (Sweden)

    Justin eSlawson

    2014-11-01

    Full Text Available CASK is an evolutionarily conserved scaffolding protein that has roles in many cell types. In Drosophila, loss of the entire CASK gene or just the CASK-β transcript causes a complex set of adult locomotor defects. In this study, we show that the motor initiation component of this phenotype is due to loss of CASK-β in dopaminergic neurons and can be specifically rescued by expression of CASK-β within this subset of neurons. Functional imaging demonstrates that mutation of CASK-β disrupts coupling of neuronal activity to vesicle fusion. Consistent with this, locomotor initiation can be rescued by artificially driving activity in dopaminergic neurons. The molecular mechanism underlying this role of CASK-β in dopaminergic neurons involves interaction with Hsc70-4, a molecular chaperone previously shown to regulate calcium-dependent vesicle fusion. These data suggest that there is a novel CASK-β-dependent regulatory complex in dopaminergic neurons that serves to link activity and neurotransmitter release.

  16. The safety of transport operations and transport casks for LWR and VVER spent fuel

    International Nuclear Information System (INIS)

    The title topics are discussed, covering the following items: safety as a basic requirement for customers and operators, regulations (which should be stringent following IAEA recommendations), Quality Assurance (which is compulsory following IAEA documents), wide transport experiences, the TN 12 spent fuel shipping cask, and the TN 120 transport/storage cask for WWER-440 spent fuel assemblies. (P.A.)

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

    Science.gov (United States)

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

    2016-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. HI-storm dry storage cask tip-over event structural response

    International Nuclear Information System (INIS)

    Current regulations in the United States (10CFR Part 72) allow the power reactor spent fuel and other radioactive materials associated with the spent fuel to be stored at an independent spent fuel storage installation, using a free-standing dry cask storage system, approved by the U. S. Nuclear Regulatory Commission. Even though a cask is designed to preclude tip-over during a design basis earthquake event, structural integrity of the cask is required to be evaluated for a non-mechanistic tip-over event. Additionally, a cask may experience a tip-over event at an angular impact velocity greater than during a design basis earthquake event, due to a potential deliberate act of terrorism of a jetliner impact into a cask storage facility. To understand how a cask storage system would perform at angular impact velocities greater than at an impact velocity greater than during an earthquake event, a study was undertaken to examine the behavior of one of the dry cask storage systems (HI-STORM 100) for a tip-over event at various angular impact velocities. Effects of changes in foundation stiffness on the cask responses were also examined. Behavior of the structural integrity of the HI-STORM 100 cask was examined using a finite-element method of analysis in a computer program, ANSYS/LS-DYNA. A detailed model of the foundation and the cask, including the exterior concrete overpack, the multi-purpose canister and the fuel basket with the spent-fuel, was developed for the explicit method of dynamic analysis. The analyses were performed for the cask tip-over impact on a concrete pad foundation at velocities of 1.7 radians/sec to 5.0 radians/sec. Additional analyses were performed for impact velocities of 1.7 radians/sec and 5.0 radians/sec with the foundation stiffness properties changed by ±50 percent. Results of the analyses were evaluated to understand the behavior of the cask, and relationship of the cask response to the impact velocity and the foundation stiffness. This

  20. Effectively meeting spent fuel storage needs with a family of dry storage casks

    International Nuclear Information System (INIS)

    During 1988--89, a number of nuclear utilities have announced their intent of developing supplemental spent fuel storage. These on-site facilities are to be operable by 1991--93. This paper discusses how the Castor ductile cast iron (DCI) storage casks is a tested and licensed means of meeting this fuel storage need. Since 1986, a total of 14 casks have been sold to the Virginia Power Co. (V.P.). Eight casks are now loaded and in storage at the V.P. Surry Nuclear Station. These casks are directly pool loaded and moved to a storage pad using straight forward handling operations. Once on the pad, there is no further need for cask operation or maintenance with this sealed and passive storage system

  1. The dry storage cask in interim storage facility and safeguards activity

    International Nuclear Information System (INIS)

    The Japan Atomic Power Company (JAPC) is preparing for interim storage of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel and to serve for about 50 years in RFSC. Metallic dry casks have already been used for spent fuel dry storage at Tokai No.2 power station. But, RFSC is not exactly the same as the dry storage facility in Tokai No.2 power station, for example, casks are transported out side of the reactor site and RFSC has no fuel handling system. Therefore, additional implementation of safeguards is necessary. This report introduces the design and handling of metallic dry casks for RFSC and the currently developing status of the safeguards activity such as containment and surveillance for the fuel loading at the power station, the cask receipt and storage at RFSC, etc. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Impact of an exploding LPG rail tank car onto a CASTOR spent fuel cask

    International Nuclear Information System (INIS)

    On 27 April 1999 a fire test was performed with a 45 m3 rail tank car partially filled with 10 m3 pressurised liquid propane. A CASTOR THTR/AVR spent fuel transport cask was positioned beside the propane tank as to suffer maximum damage from any explosion. About 17 min after fire ignition the propane tank ruptured. This resulted in a BLEVE with an expanding fireball, heat radiation, explosion overpressure, and tank fragments projected towards the cask. This imposed severe mechanical and thermal impacts directly onto the CASTOR cask, moving it 17 m from its original position. This involved rotation of the cask with the lid end travelling 10 m before it crashed into the ground. Post-test investigations of the CASTOR cask demonstrated that no loss of leaktightness or containment and shielding integrity occurred. (author)

  4. Seismic Response Analysis of Spent Nuclear Fuel Metal Storage Cask considering Soil- Structure Interaction Effects

    Energy Technology Data Exchange (ETDEWEB)

    Baeg, Chang-Yeal; Lee, Kyung-Ho; Lee, Dae-Ki [Nuclear Engineering and Technology Institute, Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Jung, In-Su; Song, Won-Tae; Jin, Han-Uk; Kim, Jong-Soo [KONES, Seoul (Korea, Republic of)

    2008-05-15

    Maintaining of the structure safety for the metal storage cask is important to store spent nuclear fuel under a seismic events. Sliding and overturning behavior must be estimated because the metal cask systems are to be installed as free standing structures on reinforced concrete pads. This behavior can cause a serious problem in the integrity of spent nuclear fuel by the impact between neighboring casks. Also, soil condition should be considered since the cask's behavior is strongly affected by the characteristics of the base soil condition. In this study, the seismic response analysis was carried out in order to evaluate the behavior of the metal storage cask under earthquake envelopment considering Soil-Structure Interaction (SSI) effects.

  5. Heat transfer tests by slice model of high performance spent fuel shipping cask

    International Nuclear Information System (INIS)

    In consideration of high burn-up plan for LWR fuel, the HP-CASK has been designed to transport more contents than the existing casks. The high performance cask required a high heat transfer performance inside the cask body, because the strong intensity of the neutron source due to the contents required use of a thick resin shielding layer. This led to a design in which internal fins welded at both ends were provided in the resin layer to ensure heat transfer performance, which was verified by means of a slice model heat transfer test. It was shown, as a result, that the heat transfer performance in the cask body had a performance as originally designed. (J.P.N.)

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

    International Nuclear Information System (INIS)

    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

  7. Selection sorting Algorithm Visualization Using Flash

    Directory of Open Access Journals (Sweden)

    Hadi Sutopo

    2011-02-01

    Full Text Available This paper is intended to develop an algorithm visualization, particularly selection sorting for an Algorithm and Programming course. Algorithm visualization technology graphically illustrates howalgorithms work. This visualization can be used to explain how all data move to the proper position in order to be sorted in a display computer for education. This research consists of 6 steps which areconcept, design, obtaining content material, assembly, testing, and distribution. During the testing step, the application is run and checked to confirm that it performs exactly what the author has intended and the students can learn selection sorting algorithm by studying the visualization. Subjects of the research were students at Department of Informatics Universitas Persada Indonesia YAI for implementation of the learning. The data were analysed using the analytic descriptive method and interpreted in a narrativeway based on the research findings. The algorithm visualization indicates that students increase their motivation and ability to program variety of sorting in programming language they learn.

  8. Card Sorts, State Tests, and Meaningful Mathematics

    Science.gov (United States)

    Chauvot, Jennifer B.; Benson, Sharon L. D.

    2008-01-01

    This article shares card-sorting activities that involve state-mandated test items to use with prospective and practicing mathematics teachers to teach about accountability measures while exploring reform-minded mathematics instruction recommendations. (Contains 2 figures.)

  9. Quantum Database Search can do without Sorting

    CERN Document Server

    Patel, A

    2001-01-01

    Sorting is a fundamental computational process, which facilitates subsequent searching of a database. It can be thought of as factorisation of the search process. The location of a desired item in a sorted database can be found by classical queries that inspect one letter of the label at a time. For an unsorted database, no such classical quick search algorithm is available. If the database permits quantum queries, however, then mere digitisation is sufficient for efficient search. Sorting becomes redundant with the quantum superposition of states. A quantum algorithm is written down which locates the desired item in an unsorted database a factor of two faster than the best classical algorithm can in a sorted database. This algorithm has close resemblance to the assembly process in DNA replication.

  10. Sorting pathways of mitochondrial inner membrane proteins

    OpenAIRE

    Mahlke, Kerstin; Pfanner, Nikolaus; Martin, Jörg; Horwich, Arthur; Hartl, Franz-Ulrich; Neupert, Walter

    1990-01-01

    Two distinct pathways of sorting and assembly of nuclear-encoded mitochondrial inner membrane proteins are described. In the first pathway, precursor proteins that carry amino-terminal targeting signals are initially translocated via contact sites between both mitochondrial membranes into the mitochondrial matrix. They become proteolytically processed, interact with the 60-kDa heat-shock protein hsp60 in the matrix and are retranslocated to the inner membrane. The sorting of subunit 9 of Neur...

  11. Another Definition of Order—Sorted Algebra

    Institute of Scientific and Technical Information of China (English)

    何自强

    1998-01-01

    In this paper the definition of order-sorted algebra is generalized by introducing transformation functions between subtypes and supertypes.According to our definition,a type needn't be a subset of its supertype and a record model may form an order-sorted algebra.A new definition of equation is given.It has also been proved that equational theories and describing single inheritance have the initial model.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    International Nuclear Information System (INIS)

    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

  14. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

  15. A nonparametric Bayesian alternative to spike sorting

    Science.gov (United States)

    Wood, Frank; Black, Michael J.

    2013-01-01

    The analysis of extra-cellular neural recordings typically begins with careful spike sorting and all analysis of the data then rests on the correctness of the resulting spike trains. In many situations this is unproblematic as experimental and spike sorting procedures often focus on well isolated units. There is evidence in the literature, however, that errors in spike sorting can occur even with carefully collected and selected data. Additionally, chronically implanted electrodes and arrays with fixed electrodes cannot be easily adjusted to provide well isolated units. In these situations, multiple units may be recorded and the assignment of waveforms to units may be ambiguous. At the same time, analysis of such data may be both scientifically important and clinically relevant. In this paper we address this issue using a novel probabilistic model that accounts for several important sources of uncertainty and error in spike sorting. In lieu of sorting neural data to produce a single best spike train, we estimate a probabilistic model of spike trains given the observed data. We show how such a distribution over spike sortings can support standard neuroscientific questions while providing a representation of uncertainty in the analysis. As a representative illustration of the approach, we analyzed primary motor cortical tuning with respect to hand movement in data recorded with a chronic multi-electrode array in non-human primates. We found that the probabilistic analysis generally agrees with human sorters but suggests the presence of tuned units not detected by humans. PMID:18602697

  16. The NCS 45 cask family: an updated design replaces an old design. Lessons learned during design, testing and licensing

    International Nuclear Information System (INIS)

    The NCS 45 cask family is intended to replace the cask types R52, TN6/1 and TN6/3. These packagings - country of origin France - were in operation worldwide since mid 1970. In the late nineties prolongations of the certificates of package approval became more and more difficult and time consuming. Finally only special arrangements for restricted contents were issued by the competent French authority which caused considerable problems when validations in other countries were applied for. To guarantee the availability of such a cask in the future for its customers NCS decided to replace the old casks by an updated design, the NCS 45 cask family

  17. Hash sort: A linear time complexity multiple-dimensional sort algorithm

    OpenAIRE

    Gilreath, William F.

    2004-01-01

    Sorting and hashing are two completely different concepts in computer science, and appear mutually exclusive to one another. Hashing is a search method using the data as a key to map to the location within memory, and is used for rapid storage and retrieval. Sorting is a process of organizing data from a random permutation into an ordered arrangement, and is a common activity performed frequently in a variety of applications. Almost all conventional sorting algorithms work by comparison, and ...

  18. Spent fuel storage and transport cask decontamination and modification. An overview of management requirements and applications based on practical experience

    International Nuclear Information System (INIS)

    A large increase in the number of casks required for transport and/or storage of spent fuel is forecast into the next century. The principal requirement will be for increased number of storage and dual purpose (transport/storage) casks for interim storage of spent fuel prior to reprocessing or permanent disposal in both on-site and off-site storage facilities. Through contact with radioactive materials spent fuel casks will be contaminated on both internal and external surfaces. In broad terms, cask contamination management can be defined by three components: minimisation, prevention and decontamination. This publication is a compilation of international experience with cask contamination problems and decontamination practices. The objective is to present current knowledge and experience as well as developments, trends and potential for new applications in this field. Furthermore, the report may assist in new design or modification of existing casks, cask handling systems and decontamination equipment

  19. The Role of the Clathrin Adaptor AP-1: Polarized Sorting and Beyond

    Directory of Open Access Journals (Sweden)

    Fubito Nakatsu

    2014-11-01

    Full Text Available The selective transport of proteins or lipids by vesicular transport is a fundamental process supporting cellular physiology. The budding process involves cargo sorting and vesicle formation at the donor membrane and constitutes an important process in vesicular transport. This process is particularly important for the polarized sorting in epithelial cells, in which the cargo molecules need to be selectively sorted and transported to two distinct destinations, the apical or basolateral plasma membrane. Adaptor protein (AP-1, a member of the AP complex family, which includes the ubiquitously expressed AP-1A and the epithelium-specific AP-1B, regulates polarized sorting at the trans-Golgi network and/or at the recycling endosomes. A growing body of evidence, especially from studies using model organisms and animals, demonstrates that the AP-1-mediated polarized sorting supports the development and physiology of multi-cellular units as functional organs and tissues (e.g., cell fate determination, inflammation and gut immune homeostasis. Furthermore, a possible involvement of AP-1B in the pathogenesis of human diseases, such as Crohn’s disease and cancer, is now becoming evident. These data highlight the significant contribution of AP-1 complexes to the physiology of multicellular organisms, as master regulators of polarized sorting in epithelial cells.

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

    International Nuclear Information System (INIS)

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

  1. Shielding benchmarks analysis for transport/storage casks

    International Nuclear Information System (INIS)

    The dose rate measurements of the TN 12/2, TN 28 VT and FS 65 has been used to evaluate the calculational procedures of Transnuclaire (TN). The three-dimensional (3D) Monte Carlo code TRIPOLI-3.4 which is used to optimize the shielding of TN casks, is applied to the analysis of a series of benchmarks. In the same cases the one-dimensional (1D) Sn code SN1D and the point kernel code MERCURE-V (3D) used for the more simplified calculations, are checked by the comparison with the measurements. The multi-group approximation used by the above codes, in order to reduces nuclear data, introduces errors due to the neutron cross-sections resonance treatment and the repartition of the gamma-ray spectrum (discrete) into an energy group structure. For a cask consisting of an iron shell of 250 mm of thickness, neutron dose rates can been underestimated of 50% if the resonances of the iron cross sections for high energy (above 1 MeV) are not taken into account. Also, depending on the energy group structure, gamma-ray dose rates can be over-estimated or under-estimated by the repartition of the gamma rays. The comparisons between measured and calculated dose rates are closer than 20% for the Monte Carlo calculations, 50% for the Sn calculations (1D) and a factor of 2 for the point kernel calculations. (author)

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

    International Nuclear Information System (INIS)

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

  3. CASTORR 1000/19: Development and Design of a New Transport and Storage Cask

    International Nuclear Information System (INIS)

    The design of the new transport and storage cask type CASTORR 1000/19 is presented in this paper. This cask was developed for the dry interim storage of spent VVER1000 fuel assemblies concerning the requirements of the Temelin NPP, Czech Republic. While the cask body is based on well-known ductile cast iron cask types with in-wall moderator, the basket follows a new concept. The basket is able to carry 19 fuel assemblies with a total decay heat power up to approximately 17 kW. The cask fulfils all requirements for a type B(U)F package. The main nuclear, mechanical and thermal properties of the cask are illustrated for normal conditions and for hypothetical accident scenarios during transport and storage. The main steps of the handling procedure such as loading the cask, drying the cavity and mounting the double lid system for tightness during interim storage are shown in principle. For this handling, boundary conditions at the NPP site such as dimensions, weight and the loading machine interface are considered. (authors)

  4. Implications of the Baltimore Rail Tunnel Fire for Full-Scale Testing of Shipping Casks

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 ∼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 ∼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

  7. Analysis and design of dry cask storage pads for plant hatch Isfsi

    International Nuclear Information System (INIS)

    An independent spent fuel storage installation (ISFSI) at Southern Company's Edwin I. Hatch Nuclear Plant (HNP) was completed, licensed, and put in service in the summer of 2000. Currently this dry cask on-site storage facility provides a temporary spent fuel storage for three Holtec HI-STAR 100 system casks. After re-racking and rod consolidation efforts, the HNP ISFSI was necessary to maintain a full core discharge capacity of its spent nuclear fuel pools and also to temporarily delay a need for a permanent off-site spent nuclear fuel repository. The HNP ISFSI was carried out to meet the following three main criteria established at the beginning of the HNP Spent Fuel Storage Project. These three criteria were 1) to use the general license approach which utilizes the license of the cask vendor rather than obtaining a site-specific license, 2) to select only dry cask products that are intended for dual purpose licensing, and 3) to acquire sufficient dry cask storage capacity to fully meet the plant's need. This paper describes the major steps of analysis and design of dry cask storage pads for Plant Hatch ISFSI. Results showed that HNP ISFSI met the applicable codes, regulatory and cask vendor requirements. (author)

  8. Optimization of cask for transport of radioactive material under impact loading

    International Nuclear Information System (INIS)

    Highlights: • Cost and weight are important criteria for fabrication and transportation of cask used for transportation of radioactive material. • Reduction of cask cost by modifying few cask geometry parameters using complex search method. • Maximum von Mises stress generated and deformation after impact as design constraints. • Up to 6.9% reduction in cost and 4.6% reduction in weight observed in the examples used. - Abstract: Casks used for transporting radioactive material need to be certified fit by subjecting them to a specific set of tests (IAEA, 2012). The high cost of these casks gives rise to the need for optimizing them. Conducting actual experiments for the process of design iterations is very costly. This work outlines a procedure for optimizing Type B(U) casks through simulations of the 9 m drop test conducted in ABAQUS®. Standard designs and material properties were chosen, thus making the process as realistic as reasonable even at the cost of reducing the options (design variables) available for optimization. The results, repeated for different source cavity sizes, show a scope for 6.9% reduction in cost and 4.6% reduction in weight over currently used casks

  9. Mechanical properties used for the qualification of transport casks: Prototype development and extension to serial production

    International Nuclear Information System (INIS)

    A thorough understanding of the mechanical behavior of material in a specific cask is required to properly analyze the structural response of the cask. An appropriate way to establish this understanding is through laboratory testing of cask material. The laboratory testing that was done to support the MOSAIK Drop Test Program is summarized as an example of how mechanical properties can be mapped for a prototype cask. The broad range behavior to be understood. This is necessary for the proper application of fracture mechanics, and focuses on fracture toughness as the inherent materials property which quantifies the fracture resistance of a material. The understanding established by a mechanics to a particular prototype, behavior of a prototype must be correctly associated with parameters which can be measured on production casks. Since the production casks cannot be destructively tested, measurements are commonly made on sub-size specimens. This may prevent direct measurement of valid design properties. An additional database may then be required to establish the correlation between sub-size specimen measurements and valid design properties. This is illustrated by outlining the additional testing which would be necessary to allow the successful verification of the MOSAIK Drop Test Program to be extended from the prototype to serially produced casks

  10. Implications of the Baltimore Rail Tunnel Fire for Full-Scale Testing of Shipping Casks

    International Nuclear Information System (INIS)

    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

  11. Two decades of experience with more than 750 CASTOR and CONSTOR transport and storage casks

    International Nuclear Information System (INIS)

    In 1983 the world-wide first dual purpose transport and storage cask - a CASTOR registered Ic-DIORIT - was loaded in Wuerenlingen/ Switzerland. Meanwhile CASTOR registered casks are used at 24 sites on four continents. Spent fuel assemblies of PWR, BWR, VVER, RBMK, FBR, MTR and THTR as well as vitrified high active waste canisters are transported and/or stored in these kinds of monolithic metal casks. MOX spent fuel of PWR and BWR has been loaded, too. Starting in the mid of the 90s, GNB developed the new CONSTOR registered cask concept, which is based on a double liner technology with a layer of heavy concrete as shielding material inbetween. This CONSTOR registered cask concept fulfils all design criteria for transport and for storage given by the IAEA recommendations and by national authorities. Up to now, more than 750 CASTOR registered and CONSTOR registered casks have been used for transports or/and loaded for longterm interim storage. More than two decades of storage experience attest to the excellent behavior of the casks including the metallic gaskets and the tightness monitoring system. Detailed measurements of temperatures and of gamma and neutron dose rates have shown in each case that the safety requirements have been fulfilled. These measurements allowed to reduce unnecessary safety margins to optimize the benefit for the user

  12. Drop accident analyses of dry metal cask without impact limiter and evaluation of leak rate

    International Nuclear Information System (INIS)

    In Japan, utilities are preparing to commence a private Interim Storage Facility (ISF) of spent fuels in Mutucity, Aomori prefecture from 2010. In the ISF, metal casks for transport and storage will be adopted and handled without an impact limiter. Cask drop tests without the impact limiter using an actual size simulated cask were carried out by CRIEPI (Central Research Institute of Electric Power Industry) in 2005. Then cases of cask drop tests were analyzed and the leak rate characteristics of a metal gasket were investigated. A general non-linear dynamic simulation computer code LS-DYNA was used in analyses. The collision velocity of the cask was calculated assuming free drop from an initial position for both horizontal drop and rotational drop. Although the drop height was 1 m in the tests, it was changed to 1.5 m and 2.0 m as parameters in the calculation for investigation of the leak rate characteristic. It was supposed that the increase of the leak rate was not only due to an increase of the total sliding movement of the lid but also caused by plastic deformation of flange or bolts. A correlation curve between total sliding movement of lid and leak rate was settled for leak rate of cask drops without the impact limier, based on results of the previous test using small-scale sized model (small scale test). Under these postulations, the leak rate could be evaluated by the correlation curve and obtained total sliding movement of the lid. In the simulated cask used for the test, a clearance between the lid and the cask body was small and the total sliding movement was limited. The leak rate estimation methodology would be applicable to the actual cask drop accident without the impact limiter, if the plastic deformation were not occurred at the flange. (author)

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

    International Nuclear Information System (INIS)

    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%

  14. Experience in complying with quality assurance requirements for cask lifting devices

    International Nuclear Information System (INIS)

    The Nuclear Assurance Corporation (NAC) owns and operates four NAC-1 truck casks. These casks are used to ship spent reactor fuel assemblies and radioactive reactor-core components. The casks have been loaded or unloaded at a total of fifteen nuclear facilities in the United States. In addition, NAC has used another large, overweight-truck cask to ship radioactive reactor core components from a reactor to a waste burial site. There are many individual differences in the cask handling facilities at each of the reactor stations, nuclear research facilities and the storage and burial sites serviced. Various types of auxiliary lifting and handling devices for on-site cask operations have been required. The quality assurance requirements for the equipment used in interfacing casks with nuclear power plant facilities have become more stringent. This paper presents details on the type of special equipment being employed, the quality assurance requirements that are imposed, and the quality assurance audits that are being performed. The paper presents NAC's experiences in the development and procurement of a variety of cask-facility equipment and the implementation of quality assurance procedures for the design, manufacture, acceptance and in-service inspection and test of the equipment. Also, experiences in working with customers' engineering and quality assurance organizations are discussed with specific attention given to the establishment of interface equipment requirements and the documentation that must be developed. The paper discusses a number of factors that must be considered in the development of design criteria for cask lifting devices. In addition to the criteria that are important to the functional safety of the equipment, other considerations important to the equipment utilization and effectiveness are presented

  15. Oxide inspecting/sorting concepts. Revision 2

    International Nuclear Information System (INIS)

    The purpose of this document is to summarize the preferred methods for inspecting and sorting plutonium and uranium oxides in preparation for future processing. A limited number of commercially available systems were investigated in preparation for selecting the preferred candidate(s). A complete listing and description of all the oxides to be processed can be located in ''Materials Disposition Acceptance Specifications for the Plutonium Immobilization Project'', document number PIP-98-047. For the purposes of this document, they will be referred to simply as oxides, unless there is a specific characteristic requiring further explanation. The physical transfer of the oxides from a convenience can into a standard oxide can will occur in the High Contamination section of the Unpackaging/Sorting glovebox. The degree of oxide inspecting and sorting performed will depend on the processing that will occur after Unpackaging/Sorting, on the known condition of the oxide when it is first received and the degree of confidence in its condition. The order in which the steps are performed and what occurs in the individual steps may vary depending on the method selected. For the purpose of organization and to help clarify what may occur in the individual steps, the following guidelines are proposed for the Unpackaging/Sorting glovebox. Sorting will involve the capture and removal of foreign particles and/or to detain hard lumps for additional processing in Crush and Grind. Part of the sorting stage will involve reducing workable lumps (de-lumping) into a smaller particle size in preparation for the sizing stage. Sizing will involve classifying/grading the oxide powder into a particle size that is acceptable for the next processing stage. It could involve some mild form of processing (screening, eTc.), but is not expected to include any sophisticated type of processing. Inspection refers to the inspection of the final product to ensure it meets the particle size requirements for

  16. Sorting by Restricted-Length-Weighted Reversals

    Institute of Scientific and Technical Information of China (English)

    Thach Cam Nguyen; Hieu Trung Ngo; Nguyen Bao Nguyen

    2005-01-01

    Classical sorting by reversals uses the unit-cost model, that is, each reversal consumes an equal cost. This model limits the biological meaning of sorting by reversal.Bender and his colleagues extended it by assigning a cost function f(l) = lα for all α≥ 0, where l is the length of the reversed subsequence. In this paper, we extend their results by considering a model in which long reversals are prohibited. Using the same cost function above for permitted reversals, we present tight or nearly tight bounds for the worst-case cost of sorting by reversals. Then we develop algorithms to approximate the optimal cost to sort a given 0/1 sequence as well as a given permutation. Our proposed problems are more biologically meaningful and more algorithmically general and challenging than the problem considered by Bender et al. Furthermore, our bounds are tight and nearly tight, whereas our algorithms provide good approximation ratios compared to the optimal cost to sort 0/1 sequences or permutations by reversals.

  17. Operation and maintenance of spent fuel storage and transportation casks/containers

    International Nuclear Information System (INIS)

    Member States have a growing need for casks for spent fuel storage and transportation. A variety of casks has been developed and is in use at an increasing number of sites. This has resulted in an accumulation of experience that will provide valuable information for other projects in spent fuel management. This publication provides a comprehensive review of information on the cask operation and maintenance associated with spent fuel storage. It draws upon generic knowledge from industrial experience and applications and is intended to serve as a basis for better planning and implementation in future projects

  18. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  19. Burnup credit application in criticality analysis of storage casks with spent RBMK-1500 nuclear fuel

    International Nuclear Information System (INIS)

    Nuclear criticality safety analysis of two types of the casks CASTOR RBMK-1500 and CONSTOR RBMK-1500 was performed using the SCALE 4.3 computer code system. These casks are planned for an interim dry storage of spent nuclear fuel at Ignalina nuclear power plant. Effective neutron multiplication factor keff was calculated for different density of the water inside the casks for unfavorable operational cases and for assumed hypothetical accident conditions when fuel in the system is fresh and fuel is depleted (i.e. burnup credit taken into account). Results show that for all cases effective neutron multiplication factor keff is less then allowable value 0.95. (author)

  20. Postmortem metallurgical examination of a fire-exposed spent fuel shipping cask

    International Nuclear Information System (INIS)

    A potmortem examination of a large fire-exposed rail-transported spent fuel shipping container has revealed the presence of two macrofissures in the outer cask shell. The first, a part-thru crack located within the seam weld fusion zone of the outer cask shell, was typical of hot cracks that may be found in stainless steel weldments. The second, located within the stainless steel base metal, apparently originated at microcracks formed during the welding of a copper-stainless steel dissimilar metal joint. The latter microcrack then propagated during the fire-test, ultimately penetrating the outer shall of the cask. 18 figures, 2 tables

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

    International Nuclear Information System (INIS)

    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

  2. Enhancing and Optimization Sorting Algorithms: An Empirical Study

    OpenAIRE

    KARIMIZADEH, Mohammad Mehdi; RAFEAZADEH, Ehsan; AMIRI, Pouran; KHOLGHNIK, Dariush

    2015-01-01

    Abstract. Sorting algorithms are used to sort a list of data. Also sorting is used in other computer operations such as searching, merging, and normalization. Since the sorting is considered as a one of the key operation in computer science, recognition of an optimization approaches can develop this science considerably. Optimization in the sorting algorithms, even in small scale, can cause saving a lot of time.  The main discussion of the paper is on those algorithms which present optimized ...

  3. Development of a Prototype Automated Sorting System for Plastic Recycling

    OpenAIRE

    D. A. Wahab; Hussain, A.; Scavino, E.; Mustafa, M.M.; Basri, H.

    2006-01-01

    Automated sorting for plastic recyclables has been seen as the way forward in the plastic recycling industry. Automated sorting provides significant improvements in terms of efficiency and consistency in the sorting process. In the case of macro sorting, which is the most common type of automated sorting, efficiency is determined by the mechanical details of the material handling system as well as the detection system. This paper provides a review on the state of-the-art technologies that hav...

  4. Software information sorting code 'PLUTO-R'

    International Nuclear Information System (INIS)

    A software information sorting code PLUTO-R is developed as one of the supporting codes of the TRITON system for the fusion plasma analysis. The objective of the PLUTO-R code is to sort reference materials of the codes in the TRITON code system. The easiness in the registration of information is especially pursued. As experience and skill in the data registration are not required, this code is usable for construction of general small-scale information system. This report gives an overall description and the user's manual of the PLUTO-R code. (author)

  5. Sorting with Asymmetric Read and Write Costs

    OpenAIRE

    Blelloch, Guy E.; Fineman, Jeremy T.; Gibbons, Phillip B.; Gu, Yan; Shun, Julian

    2016-01-01

    Emerging memory technologies have a significant gap between the cost, both in time and in energy, of writing to memory versus reading from memory. In this paper we present models and algorithms that account for this difference, with a focus on write-efficient sorting algorithms. First, we consider the PRAM model with asymmetric write cost, and show that sorting can be performed in $O\\left(n\\right)$ writes, $O\\left(n \\log n\\right)$ reads, and logarithmic depth (parallel time). Next, we conside...

  6. Radionuclide methods of sorting materials on conveyors

    International Nuclear Information System (INIS)

    Radionuclide methods suitable for sorting multi-component lump materials transported on conveyors are discussed. The methods considered use the different interaction of gamma- and X-radiations with materials of different atomic numbers. A comparison is made between a simple absorption method, a method using simultaneous absorption of photons of two different energies and a combination of the absorption and back-scattering methods. The fields of application of these methods serving to obtain output signals for the control of mechanical or pneumatic sorting devices are outlined. (author)

  7. An optimal procedure for magnet sorting

    International Nuclear Information System (INIS)

    A new magnet sorting method for accelerators is developed. It is based on the linearized analysis of the effects of errors on accelerators. It is implementable in two steps. The first step is completely analytical in character while the second step involves the comparison of computed values with the measured error values. The whole process is repeated at most ''n'' times, where ''n'' is the number of magnets to be chosen from at a time. Simulations of the method, using Mathematica reg-sign, have been implemented for sorting the APS injector synchrotron dipoles and quadrupoles with excellent results

  8. BWR-spent fuel transport and storage with the TN trademark 9/4 and TN trademark 24BH casks

    International Nuclear Information System (INIS)

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks in ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., as Muehleberg Nuclear Power Plant owner, is involved in this process and has selected to store its spent fuel, a new high capacity dual-purpose cask, the TN trademark 24BH. For the transport in a medium size cask, COGEMA LOGISTICS has developed a new cask, the TN trademark 9/4, to replace the NTL9 cask, which performed numerous transports of BWR spent fuel in the past decades. Licensed IAEA 1996, the TN trademark 9/4 is a 40 ton transport cask, for 7 BWR high burn-up spent fuel assemblies. The spent fuel assemblies can be transferred in the ZWILAG hot cell in the TN trademark 24BH cask. The first use of these casks took place in 2003. Ten TN trademark 9/4 transports were performed, and one TN trademark 24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN trademark 24BH high capacity dual purpose cask, the TN trademark 9/4 transport cask and describe in detail their characteristics and possibilities

  9. CASK stabilizes neurexin and links it to liprin-α in a neuronal activity-dependent manner.

    Science.gov (United States)

    LaConte, Leslie E W; Chavan, Vrushali; Liang, Chen; Willis, Jeffery; Schönhense, Eva-Maria; Schoch, Susanne; Mukherjee, Konark

    2016-09-01

    CASK, a MAGUK family protein, is an essential protein present in the presynaptic compartment. CASK's cellular role is unknown, but it interacts with multiple proteins important for synapse formation and function, including neurexin, liprin-α, and Mint1. CASK phosphorylates neurexin in a divalent ion-sensitive manner, although the functional relevance of this activity is unclear. Here we find that liprin-α and Mint1 compete for direct binding to CASK, but neurexin1β eliminates this competition, and all four proteins form a complex. We describe a novel mode of interaction between liprin-α and CASK when CASK is bound to neurexin1β. We show that CASK phosphorylates neurexin, modulating the interaction of liprin-α with the CASK-neurexin1β-Mint1 complex. Thus, CASK creates a regulatory and structural link between the presynaptic adhesion molecule neurexin and active zone organizer, liprin-α. In neuronal culture, CASK appears to regulate the stability of neurexin by linking it with this multi-protein presynaptic active zone complex. PMID:27015872

  10. CASKET: a computer code system for thermal and structural analyses of radioactive material transport and/or storage cask

    International Nuclear Information System (INIS)

    A computer code system CASKET (CASK thermal and structural analyses and Evaluation code system) for the thermal and structural analyses which are indispensable for radioactive material transport and/or storage cask designs has been developed. The CASKET is a simplified computer code system to perform parametric analyses on sensitivity evaluations in designing a cask and conducting its safety analysis. Main features of the CASKET are as follow: (1) it is capable to perform impact analysis of casks with shock absorbers, (2) it is capable to perform impact analysis of casks with fins. (3) puncture analysis of casks is capable, (4) rocking analysis of casks during seismic load is capable, (5) material property data library are provided for impact analysis of casks, (6) material property data library are provided for thermal analysis of casks, (7) fin energy absorption data library are provided for impact analysis of casks with fins are and (8) not only main frame computers (OS MSP) but also work stations (OS UNIX) and personal computers (OS Windows 3.1) are available. In the paper, brief illustrations of calculation methods are presented. Some calculation results are compared with experimental ones to confirm the computer programs are useful for thermal and structural analyses. (author)

  11. CASKET: a computer code system for thermal and structural analyses of radioactive material transport and/or storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-05-01

    A computer code system CASKET (CASK thermal and structural analyses and Evaluation code system) for the thermal and structural analyses which are indispensable for radioactive material transport and/or storage cask designs has been developed. The CASKET is a simplified computer code system to perform parametric analyses on sensitivity evaluations in designing a cask and conducting its safety analysis. Main features of the CASKET are as follow: (1) it is capable to perform impact analysis of casks with shock absorbers, (2) it is capable to perform impact analysis of casks with fins. (3) puncture analysis of casks is capable, (4) rocking analysis of casks during seismic load is capable, (5) material property data library are provided for impact analysis of casks, (6) material property data library are provided for thermal analysis of casks, (7) fin energy absorption data library are provided for impact analysis of casks with fins are and (8) not only main frame computers (OS MSP) but also work stations (OS UNIX) and personal computers (OS Windows 3.1) are available. In the paper, brief illustrations of calculation methods are presented. Some calculation results are compared with experimental ones to confirm the computer programs are useful for thermal and structural analyses. (author)

  12. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Storage and transportation of spent fuel and high-level waste using dry storage casks

    International Nuclear Information System (INIS)

    This paper describes the REA 2023 dry storage cask which has been designed for on-site storage and transportation of spent fuel and high-level waste. The REA 2023 is the first domestic commercial spent fuel dry storage cask completed for the Department of Energy program for demonstration of methods to improve on site utility fuel storage capacity. A description of the operations required for on-site handling and storage is provided with illustrations and photographs of the fabricated cask. An auxiliary skid is also described which is designed for both on-site handling/storage and transportation. A description of the lifting yoke and transportation impact limiters completes the total system for storage and transportation of spent fuel and high level waste in the REA 2023 casks

  16. Regulatory body experiences in licensing and inspection of dry cask storage facilities

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission (NRC), through a rigorous licensing and inspection programme, ensures the safety and security of dry cask storage. The NRC authorizes the storage of spent fuel at an independent spent fuel storage installation (ISFSI) under two licensing options: site specific licensing and general licensing. In July 1986, the NRC issued the first site specific licence to the Surry Nuclear Power Plant in Virginia, authorizing the interim storage of spent fuel in a dry storage cask configuration. Presently, there are over 40 ISFSIs licensed by the NRC, with over 800 loaded dry casks. Current projections indicate that there will be over 50 ISFSIs by the year 2010. No releases of spent fuel dry storage cask contents or other significant safety problems from the storage systems in use today have been reported. The paper discusses the NRC's licensing and inspection experiences. (author)

  17. Drop accident analyses of dry metal cask without impact limiter and evaluation of leak rate

    International Nuclear Information System (INIS)

    In Japan, utilities are preparing to initiate an independent interim storage facility (ISF) for spent fuel at Mutusi-city in Aomori prefecture in 2010. In the ISF, dual purpose metal casks which are used for both transportation and storage will be adopted, because no direct handling of spent fuel is necessary at the ISF, thereby reducing risks. The metal cask will be handled without impact limiters in the ISF. Therefore, supposing a hypothesis cask drop accident without the limiter, cask drop tests using an actual size simulated cask were analyzed and the leak characteristics from the flange with the metal gasket were investigated. The tests were conducted without the limiter, and the conditions were a horizontal drop and rotational impact with the supporting point at a trunnion. Before the calculation of this cask drop event, based on examination of results obtained from small scale tests for seal performance of flange with aged metal gasket, a correlation curve between total sliding movement of lid and leak rate was obtained. The relation between the total sliding movement of the lid and the leak rate obtained from the cask dropping tests without the impact limiter was compared with the correlation. Considering the leak rate increase due to aging of the gasket which is assumed to be ranging from 100 to 1000, the result from the cask drop tests agreed to the correlation with a 95% confidence level. Then, a general non-linear dynamic simulation computer code, LS-DYNA was used in the calculation of the cask drop tests. In the calculation, a half of the cask, considering axial symmetry, and a concrete floor were modelled. The calculation for the horizontal drop test was initiated just before a trunnion impacts the floor. For the rotational impact test, the calculation was initiated just before the edge of the outer flange impacting the floor. The impacting velocity of the cask was calculated assuming a free drop from the original position for both horizontal drop and

  18. Regulators experiences in licensing and inspection of dry cask storage facilities

    International Nuclear Information System (INIS)

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

  19. Long term containment performance test for spent fuel transport/storage casks

    International Nuclear Information System (INIS)

    The use of transport/storage cask for spent fuel storage is considered to be rational and economical. Since the storage duration may continue for 40 years or so, the function of sealing radioactive materials in the casks must be reliable for long-term. Long-term containment test of full-scale spent fuel transport/storage cask models have been in progress since 1990 in CRIEPI, Japan. It has been 11 years since it started. The results so far demonstrate and confirm very reliable containment performance of the cask lid structure with metal gaskets. Using the test data it is predicted by Larson-Miller Parameter (LMP) method that the containment system will keep its integrity at least for 40 years. (author)

  20. Development of the GA-4 and GA-9 legal weight truck spent fuel shipping casks

    International Nuclear Information System (INIS)

    General Atomics (GA) is developing two legal-weight-truck spent-fuel shipping casks for transporting commercial reactor spent fuel. The GA-4 pressurized-water-reactor (PWR) and the GA-9 boiling-water-reactor (BWR) casks are stainless steel with non-circular cross sections. Depleted uranium (DU) and boron polypropylene are used for gamma and neutron shielding. Solid pellets of boron carbide contained in a removable stainless steel fuel support structure provide criticality control. The GA-4 Cask utilizes burnup credit to maintain a capacity of four spent fuel assemblies for enrichments greater than 3.0 U-235 wt %. Aluminum honeycomb impact limiters and a dedicated semitrailer contribute to the overall efficiency and safety of the system. Design verification testing of a half-scale model cask will confirm the adequacy of the structural design

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

    International Nuclear Information System (INIS)

    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

  2. Neutronics and dose calculation for prospective spent nuclear fuel cask for Ghana Research Reactor - 1 facility

    International Nuclear Information System (INIS)

    Ghana Research Reactor-1 core is to be converted from highly enrich Uranium (HEU) fuel to low enriched Uranium (LEU) fuel in the near future: a storage cask will be needed to store the HEU fuel. Notwithstanding the core conversion process, It is also important for the facilitv to have a storage cask ready when the fuel is finally spent to temporarily store the fuel until permanent storage is provided. Winfrith Improved Multigroup Scheme-Argonne National Laboratory (WIMS-ANL). Reactor Burnup System (REBUS). Oak Ridge Isotope Generation (ORIGEN2) and Monte Carlo ''N'' Particle (MCNP5) codes have been used to design the cask. WIMS-ANL was used in generating cross sections for the REBUS code which was used in the burnup calculations. The REBUS code was used to estimate the core life time. An estimated core life of approximatcly 750 full-power-equivaicnt-days was obtained for reactor operation of 2hours a day. 4 days a week and 48 weeks in a year. The ORIGIN2 code recorded U-235 burnup weight percent of 2.90% whilst the result from the REBUS3 code was 2.86%. The amount of Pu-239 at the end of the irradiation period was 145 mg which is very low relative to other low power reactors. Isotopic inventory obtained from the ORIGIN2 and REBUS3 runs were used in setting up the MCNP5 input deck for the MCNP5 calculation of the criticality and dose rate. Six cask design options were investigated. The materials for the casks designs were selected based on their attenuation coefficient properties and their high removal cross section properties. The various materials were arranged in no specific order in multilayered casks. The reason for investigating six casks was to look at various arrangements of the cask layers that will optimize effective shielding. The spent nuclear fuel at discharge was used as the radioactivity source during the MCNP simulation. The multilayer cask shield comprise of serpentine concrete of density 5.14 g/cm3 and thickness 21.94cm which

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-11-01

    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. ANALISIS ALGORITMA INSERTION SORT, MERGE SORT DAN IMPLEMENTASINYA DALAM BAHASA PEMROGRAMAN C++

    Directory of Open Access Journals (Sweden)

    Arief Hendra Saptadi

    2013-07-01

    Full Text Available Makalah ini mengetengahkan kajian implementasi dan performa proses pengurutan menggunakan dua algoritma yang berbeda, yaitu Insertion Sort dan Merge Sort. Pada tahap pertama, kedua algoritma tersebut diimplementasikan dalam bahasa C++ untuk mengurutkan sejumlah angka yang diketikkan oleh pengguna. Pada tahap kedua, kode sumber untuk kedua algoritma tersebut diubah untuk dapat mengurutkan angka yang dihasilkan secara acak dengan jumlah angka sebanyak permintaan dari pengguna. Untuk mengetahui seberapa baik performa dalam mengurutkan data, maka dalam tahap terakhir, kedua algoritma tersebut mengurutkan sejumlah angka acak dengan rentang jumlah yang sudah ditentukan dan hasilnya kemudian dibandingkan. Dari eksperimen yang sudah dilakukan, algoritma merge sort telah memperlihatkan performa yang lebih baik, khususnya untuk jumlah data yang banyak (> 10000. Adapun algoritma insertion sort memiliki keuntungan dalam hal kompleksitas algoritma yang lebih rendah terutama dalam kondisi best case dan karena tidak menggunakan rutin rekursi dalam proses pengurutan, maka tidak membutuhkan ruang penyimpanan atau memori sebanyak algoritma Merge Sort. Kata kunci : Algoritma, Insertion Sort, Merge Sort, Performa, Bahasa C++

  6. Glucocorticoid-Dependent Complementation of a Hepatoma Cell Variant Defective in Viral Glycoprotein Sorting

    Science.gov (United States)

    John, Nancy J.; Bravo, Deborah A.; Haffar, Omar K.; Firestone, Gary L.

    1988-02-01

    We have utilized the rat hepatoma (HTC) cell sorting variant CR4 to examine the glucocorticoid-regulated pathways that localize mouse mammary tumor virus glycoproteins to the cell surface. The defective sorting of cell surface mouse mammary tumor virus glycoproteins in CR4 cells was complemented after fusion with either normal rat hepatocytes or uninfected HTC cells. Indirect immunofluorescence of transient heterokaryons revealed that the regulated localization of mouse mammary tumor virus glycoproteins was dependent upon glucocorticoid treatment and required de novo RNA and protein synthesis. Thus, a glucocorticoid-regulated trafficking activity, unrelated to mouse mammary tumor virus sequences, which is induced in both adult rat liver and cultured hepatoma cells, can act in trans to mediate an intracellular sorting pathway for membrane glycoproteins.

  7. COBRA-SFS modifications and cask model optimization

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Marginal overweight operating scenario for DOE's initiative I highway casks

    International Nuclear Information System (INIS)

    This paper assesses the potential transport of high-capacity Initiative I highway casks under development by the Office of Civilian Radioactive Waste Management (OCRWM) as permitted marginal overweight shipments that: exceed a gross vehicle weight (gvw) limit of 80,000, but weight less than 96,000 pounds; follow axle and axle group weight limits adopted by the Surface Transportation Assistance Act (STAA) of 1982; conform to dimensional restrictions to operate on most major highways; and comply with the Federal Bridge Formula. The marginal overweight tractor-trailer would operate in normal open-quotes over-the-roadclose quotes mode and comply with all laws and regulations. The vehicle would have a sleeper berth and two drivers - one to drive while the other provides escort and communications services and accumulates required off-duty time

  9. Development of design and safety analysis supporting system for casks

    International Nuclear Information System (INIS)

    Mitsubishi heavy Industries has developed a design and safety analysis supporting system 'CADDIE' (Cask Computer Aided Design, Drawing and Integrated Evaluation System), with the following objectives: (1) Enhancement of efficiency of the design and safety analysis (2) Further advancement of design quality (3) Response to the diversification of design requirements. The features of this system are as follows: (1) The analysis model data common to analyses is established, and it is prepared automatically from the model made by CAD. (2) The input data for the analysis code is available by simple operation of conversation type from the analysis model data. (3) The analysis results are drawn out in diagrams by output generator, so as to facilitate easy observation. (4) The data of material properties, fuel assembly data, etc. required for the analyses are made available as a data base. (J.P.N.)

  10. Effect of Loading Pattern on Thermal and Shielding Performance of a Spent Fuel Cask

    International Nuclear Information System (INIS)

    This study analyzes the effect of non.-uniform load patterns on peak fuel cladding temperatures and cask surface dose rates using previously validated analytical methods. The study was performed using a spent fuel storage cask that was designed to hold 24 spent fuel assemblies with a decay heat load of 24 kW. The fuel was selected to have cooling times of 3.5 to 10 years, burnups of 20 to 60 GWd/MTU, and enrichments of 2.4 to 4.8%. Three radial power distributions were considered in the study: uniform loading, hotter assemblies in the center of the cask, and hotter assemblies near the wall of the cask. Seventeen different load patterns were selected. For a given decay heat load in the cask, loading assemblies with higher decay heat output around the outside of the cask results in lower peak fuel cladding temperatures than loading hotter assemblies in the center of the cask. Several of the load patterns resulted in a peak cladding temperature that was lower than for a uniformly loaded cask. Seven source terms were selected to provide the thermal output used in the thermal analysis. A constant power density of 32 MW/MTU was used for all irradiation calculations. Cooling times were selected to provide the decay heat values used in the thermal analysis. Photon dose rates are dominated by the cobalt-60 in the bottom-end fittings, top-end fittings, and plenum and are proportional to fuel burnup. For short cooling times, photon dose rates on the side of the cask are somewhat higher due to short-lived fission products. Cask loadings with high decay heat assemblies near the periphery exhibit increased photon dose rates on the side surface and top and bottom surfaces away from the centerline. Near the centerline, on the top and bottom of the cask, the dose rates are reduced substantially. Neutron dose rates increase exponentially with burnup and are nearly independent of cooling time.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Kaushik [ORNL; Scaglione, John M [ORNL

    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.

  13. Sensitivity analyses of seismic behavior of spent fuel dry cask storage systems

    International Nuclear Information System (INIS)

    Sandia National Laboratories is conducting a research project to develop a comprehensive methodology for evaluating the seismic behavior of spent fuel dry cask storage systems (DCSS) for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission (NRC). A typical Independent Spent Fuel Storage Installation (ISFSI) consists of arrays of free-standing storage casks resting on concrete pads. In the safety review process of these cask systems, their seismically induced horizontal displacements and angular rotations must be quantified to determine whether casks will overturn or neighboring casks will collide during a seismic event. The ABAQUS/Explicit code is used to analyze three-dimensional coupled finite element models consisting of three submodels, which are a cylindrical cask or a rectangular module, a flexible concrete pad, and an underlying soil foundation. The coupled model includes two sets of contact surfaces between the submodels with prescribed coefficients of friction. The seismic event is described by one vertical and two horizontal components of statistically independent seismic acceleration time histories. A deconvolution procedure is used to adjust the amplitudes and frequency contents of these three-component reference surface motions before applying them simultaneously at the soil foundation base. The research project focused on examining the dynamic and nonlinear seismic behavior of the coupled model of free-standing DCSS including soil-structure interaction effects. This paper presents a subset of analysis results for a series of parametric analyses. Input variables in the parametric analyses include: designs of the cask/module, time histories of the seismic accelerations, coefficients of friction at the cask/pad interface, and material properties of the soil foundation. In subsequent research, the analysis results will be compiled and presented in nomograms to highlight the sensitivity of seismic response of DCSS to

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-05-01

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

  15. Evaluation of computer programs used for structural analyses of impact response of spent fuel shipping casks

    International Nuclear Information System (INIS)

    This report presents the results of a study of impact analyses of a generic spent-fuel cask. The study compares the use and results of three different finite element computer codes. Seven different cask-like model analyses are considered. The models encompass both linear and nonlinear geometric and material behavior. On the basis of the analyses results, this report recommends what parameters are useful in the comparison of different structural finite element computer programs. 5 references, 36 figures, 11 tables

  16. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Status of cask procurement strategy to satisfy DOE/OCRWM requirements

    International Nuclear Information System (INIS)

    The Nuclear Waste Policy Act requires the development of a safe and efficient system to transport spent nuclear fuel to and within the Federal Waste Management System. This paper describes the DOE/OCRWM strategy to develop and procure a major component of the Transportation System-the transport cask systems. The original initiative to develop high-capacity innovative designs and its current status is described. The follow-on phase to design and procure proven technology cask systems is also discussed

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    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.

  3. SAS1 and SAS4, two new shielding analysis sequences for spent fuel casks

    International Nuclear Information System (INIS)

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

  4. System for optical sorting of microscopic objects

    DEFF Research Database (Denmark)

    2014-01-01

    The present invention relates to a system for optical sorting of microscopic objects and corresponding method. An optical detection system (52) is capable of determining the positions of said first and/or said second objects. One or more force transfer units (200, 205, 210, 215) are placed in a...

  5. CSP description of some parallel sorting algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Linck, M.H.

    1982-11-01

    Hoare's CSP notation is used to describe 3 parallel sorting algorithms. The first algorithm uses n/2 processes working in parallel, the second uses an array of n parallel processes and the third algorithm is a parallel version of quicksort. 12 references.

  6. SELECTION SORTING ALGORITHM VISUALIZATION USING FLASH

    Directory of Open Access Journals (Sweden)

    Hadi Sutopo

    2011-02-01

    Full Text Available This paper is intended to develop an algorithm visualization, particularly selection sorting for an Algorithm and Programming course. Algorithm visualization technology graphically illustrates how algorithms work. This visualization can be used to explain how all data move to the proper position in order to be sorted in a display computer for education. This research consists of 6 steps which are concept, design, obtaining content material, assembly, testing, and distribution. During the testing step, the application is run and checked to confirm that it performs exactly what the author has intended and the students can learn selection sorting algorithm by studying the visualization. Subjects of the research were students at Department of Informatics Universitas Persada Indonesia YAI for implementation of the learning. The data were analysed using the analytic descriptive method and interpreted in a narrative way based on the research findings. The algorithm visualization indicates that students increase their motivation and ability to program variety of sorting in programming language they learn.

  7. Integration through a Card-Sort Activity

    Science.gov (United States)

    Green, Kris; Ricca, Bernard P.

    2015-01-01

    Learning to compute integrals via the various techniques of integration (e.g., integration by parts, partial fractions, etc.) is difficult for many students. Here, we look at how students in a college level Calculus II course develop the ability to categorize integrals and the difficulties they encounter using a card sort-resort activity. Analysis…

  8. Credit Scores, Race, and Residential Sorting

    Science.gov (United States)

    Nelson, Ashlyn Aiko

    2010-01-01

    Credit scores have a profound impact on home purchasing power and mortgage pricing, yet little is known about how credit scores influence households' residential location decisions. This study estimates the effects of credit scores on residential sorting behavior using a novel mortgage industry data set combining household demographic, credit, and…

  9. Gymnasielærerprofessionens sorte boks

    DEFF Research Database (Denmark)

    Lading, Aase

    eleverne destabiliserer idealiserede indre billeder af den gode lærer, og at flertydige erfaringer med lærer-elevrelationen i forlængelse deraf helst gemmes og glemmes i professionens sorte boks. Dermed indskrænkes mulighederne for professionel refleksion og diskussion af denne vigtige relations betydning...

  10. Systematic Sorting: Teacher Characteristics and Class Assignments

    Science.gov (United States)

    Kalogrides, Demetra; Loeb, Susanna; Beteille, Tara

    2013-01-01

    Although prior research has documented differences in the distribution of teacher characteristics across schools serving different student populations, few studies have examined the extent to which teacher sorting occurs within schools. This study uses data from one large urban school district and compares the class assignments of teachers who…

  11. A multispectral sorting device for wheat kernels

    Science.gov (United States)

    A low-cost multispectral sorting device was constructed using three visible and three near-infrared light-emitting diodes (LED) with peak emission wavelengths of 470 nm (blue), 527 nm (green), 624 nm (red), 850 nm, 940 nm, and 1070 nm. The multispectral data were collected by rapidly (~12 kHz) blin...

  12. 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

    International Nuclear Information System (INIS)

    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

  13. Experimental determination of radiation safety of spent nuclear fuel dry storage casks CASTOR and CONSTOR

    International Nuclear Information System (INIS)

    When Ignalina NPP was built it was planned that spent nuclear fuel (SNF) will be stored at the pools for 3-5 years and after that will be transported to Russia for reprocessing or disposal. But after reestablishment of independence the situation changed totally and an urgent need arose to solve the questions related with interim storage of spent nuclear fuel in Lithuania, because storage pools were almost totally filled. Various possibilities have been analysed and finally it was decided to use dry storage technology for interim storage (up to 50 years) of Ignalina NPP spent nuclear fuel. For this purpose GNB (Germany) duel-purpose casks have been chosen. The part of them are ductile cast iron CASTOR RBMK-1500 casks and the rest part are metal-concrete CONSTOR RBMK-1500 casks. In order to evaluate radiation characteristics of the casks, combined experimental investigations (measurements of the equivalent dose and γ-spectrum on the cask surface at dry storage) and computer modeling (calculations of the equivalent dose rates, activities of nuclides, etc.) were performed. The obtained results show that equivalent dose rate values on the surface of the casks are much less than the design criteria value of 1000 μSv/h. (author)

  14. Seismic response analysis of a free-standing model of spent fuel storage cask

    International Nuclear Information System (INIS)

    The seismic response analyses of a free-standing spent fuel storage cask are performed for an artificial time history acceleration generated on the basis of the US NRC RG1.60 response acceleration spectrum. This paper focuses on the structural stability regarding seismic loads to check the overturning possibility of a storage cask and the slip displacement on the concrete installation bed. A simple structural analysis model for the storage cask is developed to perform the parametric effect analyses regarding the seismic responses. Two parameters considered in the analyses are the magnitude of the seismic load and the interface friction between the cask's bottom surface and the upper surface of the concrete installation bed. The analyses results show that the seismic responses of the storage cask are influenced by a combination of the two parameters and the storage cask also has a large marginal integrity for the maximum overturning angle and the slip distance for the design and beyond design seismic loads. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  17. Annex IV. The technical and logistic benefits of non-uniform, zoned cask loading

    International Nuclear Information System (INIS)

    The purpose of this appendix is to describe the benefits of licensing and using non-uniform, zoned loading of casks, including the physical nature of the phenomena that underlie those benefits. Based on the systematics of the zoned loading analysis sequence, this appendix also outlines a regulatory approach for licensing and specifying the range of couplings of the outer and inner zone fuel characteristics that result in total external dose rates being at the regulatory limit. This Appendix has outlined an approximate method for evaluating the capability of zone loaded casks, and has used that method to evaluate zoned loading in a typical long term shipping situation. The results of the evaluation indicate that there are two types of benefit arising from the use of zoned cask loading when coupled with an optimised long term plan for fuel selection to accomplish the loadings. A technical benefit in which the radioactivity content of a cask is increased without an increase in the external dose rate, and a logistic benefit, realised through the use of an appropriate long term fuel selection and cask-loading plan, that significantly extends the usability of a cask design, delivers shipments with characteristics that are fairly stable over time, and is consistently loaded close to its license limit

  18. Optimization strategies for cask design and container loading in long term spent fuel storage

    International Nuclear Information System (INIS)

    As delays are incurred in implementing reprocessing and in planning for geologic repositories, storage of increasing quantities of spent fuel for extended durations is becoming a growing reality. Accordingly, effective management of spent fuel continues to be a priority topic. In response, the IAEA has organized a series of meetings to identify cask loading optimisation issues in preparation for a technical publication on Optimization Strategies for Cask/Container Loading in Long Term Spent Fuel Storage. This publication outlines the optimisation process for cask design, licensing and utilization, describing three principal groups of optimization activities in terms of relevant technical considerations such as criticality, shielding, structural design, operations, maintenance and retrievability. The optimization process for cask design, licensing, and utilization is outlined. The general objectives for the design of storage casks, including storage casks that are intended to be transportable, are summarized. The nature of optimization within the design process is described. The typical regulatory and licensing process is outlined, focusing on the roles of safety regulations, the regulator, and the designer/applicant in the optimization process. Based on the foregoing, a description of the three principal groups of optimization activities is provided. The subsequent chapters of this document then describe the specific optimization activities within these three activity groups, in each of the several design disciplines

  19. Long-term containment performance of storage cask for spent fuel

    International Nuclear Information System (INIS)

    Previous papers [1-4] reported that the performance life of metallic gaskets coated with aluminum or silver will be more than 190 years, respectively, based on an accelerated tests and Larson-Miller's estimation method. This paper describes demonstrative tests on long-term containment of full-scale cask lid models. The cask models were selected from various types of storage casks, taking account of the influential structure such as lid shape, gasket groove, and gasket structure. The tests have continued from 1990 for more than 9 years. In addition, it was noted that the casks experience temperature variation of seasons in the storage building. It will be necessary to confirm any influence of such environmental temperature variation, on the containment performance of the casks. Primary, a cyclic creep characteristics should be investigated. Factors for the cyclic creep will be number of cycles, temperature, velocity of cycle, etc. Taking account of those factors, temperature-cyclic tests were carried out to investigate the effect on the containment of the cask. (author)

  20. Sorting and targeting of melanosomal membrane proteins: signals, pathways, and mechanisms.

    Science.gov (United States)

    Setaluri, V

    2000-06-01

    Newly synthesized melanosomal proteins, like many other cellular proteins, traverse through a series of intracellular compartments en route to melanosomes. Entry and exit of proteins through these compartments is orchestrated by cellular sorting machinery that recognize specific sorting signals. Melanosomal membrane proteins begin their intracellular journey upon co-translational importation into the endoplasmic reticulum (ER). The biosynthetic output of tyrosinase, the key melanogenic enzyme, appears to be regulated by quality-control events at the ER, the 'port of entry' to the secretory pathway. Following maturation in the ER and through the Golgi, the sorting of these proteins in the trans-Golgi network for intracellular retention and transport along endosome/lysosome pathway requires cytoplasmically exposed signals. A di-leucine motif, present in the cytoplasmic tails of most melanosomal proteins, and its interaction with adaptor protein (AP) complexes, specifically AP-3, are critical for these events. Defects in sorting signals and the cytosolic components that interact with these signals result in a number of murine coat color phenotypes and cause human pigmentary disorders. Thus, missense or frame-shift mutations that produce truncated tyrosinase lacking the melanosomal sorting signal(s) appear to be responsible for murine platinum coat color phenotypes and a proportion of human oculocutaneous albinism-1; mutations in AP-3 appear to be responsible for the mocha phenotype in mice and Hermansky-Pudlak-like syndrome in man. Additional signals and sorting steps downstream of AP-3 appear to be required for endosomal sorting and targeting proteins to melanosomes. Signals and mechanisms that sequester melanosomal proteins from endosomes/lysosomes are not understood. Potential candidates that mediate such processes include proteins encoded by lyst and pallid genes. The common occurrence of abnormalities in melanosomes in many storage-pool disorders suggests that

  1. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    International Nuclear Information System (INIS)

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks

  2. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, O.G.; Matveev, V.Z.; Morenko, A.I.; Il' kaev, R.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)

    2004-07-01

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks.

  3. Analysis for Spent Nuclear Fuel Multi-Canister Overpack (MCO) Drop into the Cask from the Multi-Canister Overpack - Handling Machine (MHM) with Air Cushion

    International Nuclear Information System (INIS)

    The purpose of this report is to investigate the potential for damage to the MCO during impact from an accidental drop from the MHM into the shipping cask. The MCO is dropped from a height of 8.2 feet above the cask enters the cask concentrically and falls the additional 12.83 feet to the cask bottom. Because of the interface fit between the MCO and the cask and the air entrapment the MCO fall velocity is slowed. The shipping cask is resting on an impact absorber at the time of impact. The energy absorbing properties of the impact absorber are included in this analysis

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

    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

  6. Development of a Prototype Automated Sorting System for Plastic Recycling

    Directory of Open Access Journals (Sweden)

    D. A. Wahab

    2006-01-01

    Full Text Available Automated sorting for plastic recyclables has been seen as the way forward in the plastic recycling industry. Automated sorting provides significant improvements in terms of efficiency and consistency in the sorting process. In the case of macro sorting, which is the most common type of automated sorting, efficiency is determined by the mechanical details of the material handling system as well as the detection system. This paper provides a review on the state of-the-art technologies that have been deployed by some of the recycling facilities abroad. The design and development of a cost effective prototype automated system for sorting plastic recyclables is proposed and discussed.

  7. PhySortR: a fast, flexible tool for sorting phylogenetic trees in R

    Science.gov (United States)

    Stephens, Timothy G.; Bhattacharya, Debashish; Ragan, Mark A.

    2016-01-01

    A frequent bottleneck in interpreting phylogenomic output is the need to screen often thousands of trees for features of interest, particularly robust clades of specific taxa, as evidence of monophyletic relationship and/or reticulated evolution. Here we present PhySortR, a fast, flexible R package for classifying phylogenetic trees. Unlike existing utilities, PhySortR allows for identification of both exclusive and non-exclusive clades uniting the target taxa based on tip labels (i.e., leaves) on a tree, with customisable options to assess clades within the context of the whole tree. Using simulated and empirical datasets, we demonstrate the potential and scalability of PhySortR in analysis of thousands of phylogenetic trees without a priori assumption of tree-rooting, and in yielding readily interpretable trees that unambiguously satisfy the query. PhySortR is a command-line tool that is freely available and easily automatable. PMID:27190724

  8. Smart Sort: Design and Analysis of a Fast, Efficient and Robust Comparison Based Internal Sort Algorithm

    CERN Document Server

    Singh, Niraj Kumar

    2012-01-01

    Smart Sort algorithm is a "smart" fusion of heap construction procedures (of Heap sort algorithm) into the conventional "Partition" function (of Quick sort algorithm) resulting in a robust version of Quick sort algorithm. We have also performed empirical analysis of average case behavior of our proposed algorithm along with the necessary theoretical analysis for best and worst cases. Its performance was checked against some standard probability distributions, both uniform and non-uniform, like Binomial, Poisson, Discrete & Continuous Uniform, Exponential, and Standard Normal. The analysis exhibited the desired robustness coupled with excellent performance of our algorithm. Although this paper assumes the static partition ratios, its dynamic version is expected to yield still better results.

  9. Ferrofluid mediated nanocytometry.

    Science.gov (United States)

    Kose, Ayse Rezzan; Koser, Hur

    2012-01-01

    We present a low-cost, flow-through nanocytometer that utilizes a colloidal suspension of non-functionalized magnetic nanoparticles for label-free manipulation and separation of microparticles. Our size-based separation is mediated by angular momentum transfer from magnetically excited ferrofluid particles to microparticles. The nanocytometer is capable of rapidly sorting and focusing two or more species, with up to 99% separation efficiency and a throughput of 3 × 10(4) particles/s per mm(2) of channel cross-section. The device is readily scalable and applicable to live cell sorting with biocompatible ferrofluids, offering competitive cytometer performance in a simple and inexpensive package. PMID:22076536

  10. BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

    International Nuclear Information System (INIS)

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TNTM24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TNTM9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TNTM9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TNTM24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TNTM9/4 round trips are performed, and one TNTM24BH is loaded. 5 additional TNTM24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TNTM24BH high capacity dual purpose cask and the TNTM9/4 transport cask characteristics and capabilities will then be detailed. (authors)

  11. Inspection of Fuel Cladding and Metal Gasket in Metallic Dry Cask at Tokai No. 2 Power Station

    International Nuclear Information System (INIS)

    The metallic dry cask storage of spent fuel started in December 2001 at TOKAI No.2 power station. The cask that had served for 7 years was inspected in January 2009. The objective of this inspection is confirmation of fuel cladding and metal gasket integrity. This cask accommodates 8 × 8 zirconium liner type fuel. The gasket applied to this cask consists of aluminum outer lining and Inconel spring. This inspection confirmed that there had been no damage in fuel cladding and metal gasket during the storage for 7 years. (author)

  12. Research on spent fuel storage and transportation in CRIEPI. Part 1. Metal cask and vault storages, and transportation

    International Nuclear Information System (INIS)

    For metal cask storage method, containment safety of the metal gasket was demonstrated using a full-scale metal cask without shock absorbers subjected to drop accidents during handling work at a storage facility. The instantaneous leakage was negligible. Long-term containment of the metal gasket has been demonstrated using full-scale cask lid models under a high and constant temperature for more than 17 years. Taking account of temperature decay in the real cask, the containment for more than 60 years has been evaluated. Hypothetical airplane crash against a cask storage building was studied by analysis and tests. The mechanical impact on the containments of the metal gaskets of the lid structure of the cask was analyzed and demonstrated by tests. Vault storage method may be economical for a large capacity of spent fuel storage. A design concept of the vault storage at a shallow underground was developed and the licensability of the underground's space was studied. Transport cask may deteriorate with respect to its elastomer gaskets as a result of creep deformation. The deformation and reduction of resilience of the gasket was studied by means of an analysis of finite element method. Transport ship of casks on the sea was assumed to shipwreck hypothetically and the casks loose their containment in the sea. Radiation dose under the hypothetical accident was evaluated by means of an analysis using an oceanic circulation model of the sea water. (author)

  13. A mower detector to judge soil sorting

    Energy Technology Data Exchange (ETDEWEB)

    Bramlitt, E.T.; Johnson, N.R. [Thermo Nuclear Services, Inc., Albuquerque, NM (United States)

    1995-12-31

    Thermo Nuclear Services (TNS) has developed a mower detector as an inexpensive and fast means for deciding potential value of soil sorting for cleanup. It is a shielded detector box on wheels pushed over the ground (as a person mows grass) at 30 ft/min with gamma-ray counts recorded every 0.25 sec. It mirror images detection by the TNS transportable sorter system which conveys soil at 30 ft/min and toggles a gate to send soil on separate paths based on counts. The mower detector shows if contamination is variable and suitable for sorting, and by unique calibration sources, it indicates detection sensitivity. The mower detector has been used to characterize some soil at Department of Energy sites in New Jersey and South Carolina.

  14. Gender Sorting at the Application Interface

    OpenAIRE

    Fernandez, Roberto; Friedrich, Colette

    2011-01-01

    We document gender sorting of candidates into gender-typed jobs at the point of initial application to a company. At this step of the hiring process, the firm has implemented a policy whereby organizational screeners’ discretion has been eliminated such that there is no opportunity for contact between hiring agents and applicants. Thus, the job choices studied here offer unique insight as they are uncontaminated by screeners’ steering of candidates toward gender-typed jobs. Eve...

  15. Micro-fluidic chip for cell sorting

    Czech Academy of Sciences Publication Activity Database

    Šerý, Mojmír; Pilát, Zdeněk; Ježek, Jan; Kaňka, Jan; Zemánek, Pavel

    Munich : EOS, 2015. ISBN 978-952-93-5069-8. [EOS Conferences at the World of Photonics Congress 2015. Munich (DE), 22.06.2015-25.06.2015] R&D Projects: GA MŠk(CZ) LD14069; GA MŠk(CZ) LO1212; GA TA ČR TA03010642; GA MŠk ED0017/01/01 Institutional support: RVO:68081731 Keywords : Micro-fluidic chip * cell sorting Subject RIV: BH - Optics, Masers, Lasers

  16. Loudness effect on pairwise comparisons and sorting

    OpenAIRE

    Susini, Patrick; Houix, Olivier; SAINT PIERRE, Guillaume

    2015-01-01

    Effect of loudness on the perceptual structure underlying a corpus of sounds is investigated bytwo experimental methods: pairwise comparisons and sorting. Both methods are applied to a corpus of recordingssounds presented with their ecological, non normalized loudness, and to the same corpus equalized in loudness. Twotypes of perceptual structures (multidimensional scaling and hierarchical cluster analysis) are derived. Domination ofone auditory attribute ? loudness ? on less salient ones is ...

  17. Parallel Algorithms for Neuronal Spike Sorting

    OpenAIRE

    Bergheim, Thomas Stian; Skogvold, Arve Aleksander Nymo

    2011-01-01

    Neurons communicate through electrophysiological signals, which may be recorded using electrodes inserted into living tissue.When a neuron emits a signal, it is referred to as a spike, and an electrode can detect these from multiple neurons.Neuronal spike sorting is the process of classifying the spike activity based on which neuron each spike signal is emitted from.Advances in technology have introduced better recording equipment, which allows the recording of many neurons at the same time.H...

  18. Fluorescence-Activated Nucleolus Sorting in Arabidopsis.

    Science.gov (United States)

    Pontvianne, Frédéric; Boyer-Clavel, Myriam; Sáez-Vásquez, Julio

    2016-01-01

    Nucleolar isolation allows exhaustive characterization of the nucleolar content. Centrifugation-based protocols are not adapted to isolation of nucleoli directly from a plant tissue because of copurification of cellular debris. We describe here a method that allows the purification of nucleoli using fluorescent-activated cell sorting from Arabidopsis thaliana leaves. This approach requires the expression of a specific nucleolar protein such as fibrillarin fused to green fluorescent protein in planta. PMID:27576720

  19. Sorting Network for Reversible Logic Synthesis

    CERN Document Server

    Islam, Md Saiful; Mahmud, Abdullah Al; karim, Muhammad Rezaul

    2010-01-01

    In this paper, we have introduced an algorithm to implement a sorting network for reversible logic synthesis based on swapping bit strings. The algorithm first constructs a network in terms of n*n Toffoli gates read from left to right. The number of gates in the circuit produced by our algorithm is then reduced by template matching and removing useless gates from the network. We have also compared the efficiency of the proposed method with the existing ones.

  20. Optical sorting due to optical binding

    Czech Academy of Sciences Publication Activity Database

    Karásek, Vítězslav; Zemánek, Pavel

    Bellingham: SPIE, 2013, 881027:1-8. ISSN 0277-786X. [Optical Trapping and Optical Micromanipulation /10./. San Diego (US), 25.08.2013-29.08.2013] R&D Projects: GA ČR GPP205/12/P868 Institutional support: RVO:68081731 Keywords : optical binding * optical sorting * particles * optical trapping * bessel beam s * code division multiplexing * numerical simuklations Subject RIV: BH - Optics, Masers, Laser s

  1. A sorting network in bounded arithmetic

    Czech Academy of Sciences Publication Activity Database

    Jeřábek, Emil

    2011-01-01

    Roč. 162, č. 4 (2011), s. 341-355. ISSN 0168-0072 R&D Projects: GA AV ČR IAA1019401; GA MŠk(CZ) 1M0545 Institutional research plan: CEZ:AV0Z10190503 Keywords : bounded arithmetic * sorting network * proof complexity * monotone sequent calculus Subject RIV: BA - General Mathematics Impact factor: 0.450, year: 2011 http://www.sciencedirect.com/science/article/pii/S0168007210001272

  2. Active sorting switch for biological objects

    Czech Academy of Sciences Publication Activity Database

    Šerý, Mojmír; Pilát, Zdeněk; Jonáš, Alexandr; Ježek, Jan; Jákl, Petr; Zemánek, Pavel; Samek, Ota; Nedbal, Ladislav; Trtílek, M.

    Bellingham : SPIE, 2010, 776210: 1-7. ISBN 978-0-8194-8258-7. [Optical Trapping and Optical Micromanipulation /7./. San Diego (US), 01.08.2010] R&D Projects: GA MPO FR-TI1/433; GA MŠk OC08034 Institutional research plan: CEZ:AV0Z20650511; CEZ:AV0Z60870520 Keywords : laser tweezers * laser diode * optical sorting * microfluidics Subject RIV: BH - Optics, Masers, Lasers

  3. How Schwann Cells Sort Axons: New Concepts.

    Science.gov (United States)

    Feltri, M Laura; Poitelon, Yannick; Previtali, Stefano Carlo

    2016-06-01

    Peripheral nerves contain large myelinated and small unmyelinated (Remak) fibers that perform different functions. The choice to myelinate or not is dictated to Schwann cells by the axon itself, based on the amount of neuregulin I-type III exposed on its membrane. Peripheral axons are more important in determining the final myelination fate than central axons, and the implications for this difference in Schwann cells and oligodendrocytes are discussed. Interestingly, this choice is reversible during pathology, accounting for the remarkable plasticity of Schwann cells, and contributing to the regenerative potential of the peripheral nervous system. Radial sorting is the process by which Schwann cells choose larger axons to myelinate during development. This crucial morphogenetic step is a prerequisite for myelination and for differentiation of Remak fibers, and is arrested in human diseases due to mutations in genes coding for extracellular matrix and linkage molecules. In this review we will summarize progresses made in the last years by a flurry of reverse genetic experiments in mice and fish. This work revealed novel molecules that control radial sorting, and contributed unexpected ideas to our understanding of the cellular and molecular mechanisms that control radial sorting of axons. PMID:25686621

  4. Conceptual design report for a transportable DUCRETE spent fuel storage cask system

    International Nuclear Information System (INIS)

    A conceptual design has been developed for a spent fuel dry storage cask that employs depleted uranium concrete (DUCRETE) in place of ordinary concrete. DUCRETE, which uses depleted uranium oxide rocks rather than gravel as the concrete's heavy aggregate, is a more efficient overall radiation shield (gamma and neutron) than either steel or ordinary concrete. Thus, it allows the cask weight and size to be substantially reduced. Also, using DUCRETE as shielding avoids, or at least defers, disposal of the depleted uranium as waste. This report focuses on DUCRETE cask transportation issues. The approach studied involves placing the storage cask into a simple steel transportation overpack. Preliminary analyses were performed to demonstrate the transportation system's ability to meet the structural, thermal, and shielding transportation criteria. Conservative manual calculations were performed to demonstrate the adequacy of the DUCRETE transportation overpack with respect to structural requirements. Two-dimensional thermal analyses were performed on the system (the DUCRETE storage cask inside the steel overpack) using the ANSYS thermal analysis code. Two-dimensional shielding analyses were performed on the system with the MCNP code. Effects of the fuel axial burnup profile and solar radiation are considered. The analyses show that the proposed system can meet the transportation structural criteria and can easily meet the transportation shielding criteria. The thermal criteria are not as easy to meet because when the storage cask is placed horizontally in the transportation overpack, the DUCRETE storage cask's ventilation duct becomes an insulating dead air space. The maximum allowable temperature for the DUCRETE, which is not yet known, will be the limiting factor

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  6. Sorting of Lipids and Proteins in Membrane Curvature Gradients

    OpenAIRE

    Tian, A.; Baumgart, T.

    2009-01-01

    The sorting of lipids and proteins in cellular trafficking pathways is a process of central importance in maintaining compartmentalization in eukaryotic cells. However, the mechanisms behind these sorting phenomena are currently far from being understood. Among several mechanistic suggestions, membrane curvature has been invoked as a means to segregate lipids and proteins in cellular sorting centers. To assess this hypothesis, we investigate the sorting of lipid analog dye trace components be...

  7. Radiological and thermal characteristics of CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks for spent nuclear fuel storage at Ignalina nuclear power plant

    International Nuclear Information System (INIS)

    The main characteristics, such as temperatures of the fuel rod cladding and cask surface, dose rates at the surface and at the some distance for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks loaded with spent nuclear fuel are presented. These casks are used for an interim dry storage of spent nuclear fuel at Ignalina Nuclear Power Plant. Numerical modeling (calculation of the equivalent dose rates, activities of nuclides, etc.) and experimental measurements of the equivalent dose and gamma spectrum on the cask surface at the dry storage facility were performed for assessment of radiation characteristics. Temperatures were evaluated using only numerical modeling. Rather good agreement between experimentally determined and calculated dose rates for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks was obtained. Also it was revealed that maximum fuel rod cladding temperature is higher for CONSTOR RBMK-1500 cask, but never exceeds the maximum allowable value. The cask surface temperatures are similar for both cask types. (orig.)

  8. Categorizing Variations of Student-Implemented Sorting Algorithms

    Science.gov (United States)

    Taherkhani, Ahmad; Korhonen, Ari; Malmi, Lauri

    2012-01-01

    In this study, we examined freshmen students' sorting algorithm implementations in data structures and algorithms' course in two phases: at the beginning of the course before the students received any instruction on sorting algorithms, and after taking a lecture on sorting algorithms. The analysis revealed that many students have insufficient…

  9. Applying Single Kernel Sorting Technology to Developing Scab Resistant Lines

    Science.gov (United States)

    We are using automated single-kernel near-infrared (SKNIR) spectroscopy instrumentation to sort fusarium head blight (FHB) infected kernels from healthy kernels, and to sort segregating populations by hardness to enhance the development of scab resistant hard and soft wheat varieties. We sorted 3 r...

  10. Gender Sorting across K-12 Schools in the United States

    Science.gov (United States)

    Long, Mark C.; Conger, Dylan

    2013-01-01

    This article documents evidence of nonrandom gender sorting across K-12 schools in the United States. The sorting exists among coed schools and at all grade levels, and it is highest in the secondary school grades. We observe some gender sorting across school sectors and types: for instance, males are slightly underrepresented in private schools…

  11. My eSorts and Digital Extensions of Word Study

    Science.gov (United States)

    Zucker, Tricia A.; Invernizzi, Marcia

    2008-01-01

    "My eSorts" is a strategy for helping children learn to read and spell in a socially motivated context. It is based on developmental spelling research and the word study approach to teaching phonics and spelling. "eSorting" employs digital desktop publishing tools that allow children to author their own electronic word sorts and then share these…

  12. Conceptual Design Report - Cask Loadout System Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform, 105 K West Basin, Project A.5/A.6

    International Nuclear Information System (INIS)

    This conceptual design report documents the redesign of the immersion pail support structure (IPSS) and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5lA.6, Canister Transfer Facility Modifications. Project A.5lA.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The junction 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 slablwall 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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Optimization of radiation protection by optimizing technology of CASTOR-Cask loading

    International Nuclear Information System (INIS)

    Full text: Germany Optimization of Protection is one of the basic principles of the ICRP System of Radiation Protection. Often this principle is misunderstood and people try to achieve minimal doses irrespective of the amount of manpower or money they have to afford to reach this aim. The better way of optimization is to optimize the technology or the practise which is the cause of radiation exposure and at the same time reduce the dose uptake. Three measures have been used for this purpose in the management of spent fuel in Germany in preparation for the dry storage in CASTOR-Casks. The casks have to be loaded with the spent fuel in the pond of the power plant. After the loading the cask has to be dewatered and dried. The remaining humidity has to be checked with respect to a given maximum residual humidity to avoid corrosion during the long-term storage. Initially a measuring device using the dew point mirror method was used. The mirror was often polluted and needed recalibration. This led to a large variety of measuring times, the time period needed for the above mentioned three steps ranged from 55 to 120 hours. Thus the work could not be reliably planned. To solve this problem we now use a pressure-rise method to measure the humidity within the cask. The time needed is now nearly equal and reliable for all cask loadings and considerably lower than using the dew point method. Thereby the dose uptake of the cask handling staff could be reduced to 2.5 man mSv on average in comparison to the former collective dose of 4 to 5 man mSv. A second step for reducing the dose of the staff is the introduction of remotely controlled valves for the drying process, the humidity measurement and the subsequent filling with Helium. The valves are located at the lid of the cask where a remarkable dose rate could be. The equipment for the remote valve handling has been successfully tested. In the same line is a third measure: to record the process data by computer. The supervising

  15. Evaluation of Impact Resistance of Concrete Overpack of Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    The concrete overpack of the cask provides radiation shielding as well as physical protection for inner canister against external mechanical shock. When the overpack undergoes a severe missile impact which might be caused by tornado or aircraft crash, it should sustain minimal level of structural integrity so that the radiation shielding and the retrievability of canister are maintained. Empirical formulas have been developed for the evaluation of concrete damage but those formulas can be used only for local damage evaluation and not for global damage evaluation. In this research, a series of numerical simulations and tests have been performed to evaluate the damage of two types of concrete overpack segment models under high speed missile impact. It is shown that appropriate modeling of material failure is crucial in this kind of analyses and finding the correct failure parameters may not be straightforward. When comparing the simulation results with the test results, it is shown that neither setting, case 1 and 2 provides results with consistent agreement with test results. That is, case 1 setting is more close to reality in Type 1 model analysis, but for Type 2, case 2 setting provides more close results to the reality. In both the case, not enough deformation is predicted by simulation compared to the tests. Weak failure and eroding criteria give larger penetration depth with insufficient overall damage due to energy loss with element erosion.

  16. A CFD analysis of thermal behaviour of transportation cask under fire test conditions

    International Nuclear Information System (INIS)

    Highlights: → Melting and natural convection of lead in cask has been studied using CFD for the first time. → The role of turbulent natural convection on melting was pronounced. → The study establishes the importance of natural convection for accurate thermal design of cask. - Abstract: Thermal design of transportation cask for shipping radioactive waste needs strict compliance with the guidelines of the regulatory bodies. Lead shielding is usually provided in these casks for arresting gamma rays and reducing hazardous emissions to the environment below permissible limits. During transportation, accidental fire may break out and cause melting of lead for a prescribed duration. The present analysis reports, for the first time, a comprehensive CFD analysis of the thermal behaviour of melting of lead under high Rayleigh number convection during the fire test. The study reveals a substantial influence of natural convection on the thermal state and melting behaviour of lead which may have a great bearing on the safety and security of public during transportation of cask.

  17. A study on the free drop impact of a cask using commercial FEA codes

    International Nuclear Information System (INIS)

    The package used to transport radioactive materials, which is called a cask, must be designed to keep its contents safe under normal and hypothetical accident conditions. The design requirements of the cask are verified by test or finite element analysis (FEA). Comparing evaluation procedures for the safety of a new cask, the cost of FEA is generally much less than that test. Therefore, FEA is mainly used to verify safety of a cask under the considered conditions. However, one commercial FEA code may show different results from another FEA code for the same problem due to the modeler's several assumptions for simplifying actual states into the FE model and due to modeling technique. Materials of the components of a cask display elastic-plastic or elastic-perfectly plastic behavior under the considered conditions in which large deformation, impact and contact mechanism are included. The behavior is simulated with difficulty and may have different results depending on FEA codes. In this paper, finite element analysis is carried out for the 9-m free drop and the puncture condition under the hypothetical accident condition by using LS-DYNA3D and ABAQUS/Explicit. Energy and effective stress on each component are presented and compared between the two FEA codes, where the effective stress designates the maximum von Mises stress on inner and outer shells

  18. Thermal analysis of spent fuel storage cask using the fluent code

    International Nuclear Information System (INIS)

    Thermal analysis for spent fuel storage cask loaded with 24 spent PWR fuel assemblies has been carried out using the Fluent code to verify the reliability of analysis method and procedure. And the temperature distribution for storage cask loaded with 24 metalized fuels equivalent to 96 PWR fuels has been also calculated. It is found that the storage volume of PWR assembly is reduced to a quarter and the heat load is reduced to a half by the preferential elimination of Sr-90 and Cs-137 through the metalization process of spent PWR fuel. Total decay heat from 24 spent PWR fuels and 24 metalized spent fuels are 28 kW and 54 kW, respectively. The calculated temperatures for 24 spent PWR fuels were compared with the proven data presented from the safety analysis report of spent fuel storage cask. It has good agreement between the two results, and it is also found that the feasibility of the analysis method and procedure has been confirmed by the results to estimate the temperature for the spent fuel storage cask. The maximum fuel temperature for 24 metalized spent fuel assemblies inside the cask is calculated at 617 .deg. C

  19. Integrated cask storage systems for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Since 1979 Tennessee Valley Authority TVA has participated in conceptual design studies of dry storage vaults, silos, casks, ad dry wells, and, with DOE and others, has undertaken limited demonstrations of rod consolidation and cask dry storage at TVA's Browns Ferry Nuclear Plant in Alabama. TVA believes the integrated storage cask concept is worthy of consideration as an alternative for spent fuel management. Placing spent fuel in a secure passive storage mode at an early date and avoiding unnecessary handling and repackaging reduces the potential for occupational and public radiological exposure. Therefore the notion of a universal cask used for storage, shipment, and disposal is appealing from a safety, environmental, and public perception standpoint. The universal cask can also serve as a dispersed monitored retrievable storage (MRS), thus eliminating the need for redundant facilities, and it does not foreclose future options. It also appears that this concept would simplify repository design, ease retrievability, and provide greater flexibility in repository siting. 2 figures, 2 tables

  20. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.